ML24311A121
| ML24311A121 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/19/2024 |
| From: | David Wrona Plant Licensing Branch II |
| To: | Erb D Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2024-LRO-0020 | |
| Download: ML24311A121 (8) | |
Text
December 19, 2024 Delson Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNIT 1 - EVALUTION OF EFFECTS OF OUT-OF-LIMIT CONDITION AS DESCRIBED IN ASME SECTION XI, IWB-3720(a) (EPID L-2024-LRO-0028)
Dear Delson Erb:
By letter dated June 26, 2024, Tennessee Valley Authority (TVA) submitted an evaluation of the pressure-temperature excursion experienced at Browns Ferry Nuclear Plant, Unit 1 (BFN1) following a reactor scram on April 24, 2024. TVA submitted the evaluation for U.S. Nuclear Regulatory Commission (NRC or Commission) review and approval in accordance with the Commissions requirement in section 50.55a(b)(2)(xliii) of Title 10 of the Code of Federal Regulations (10 CFR) that was in effect at that time of the submittal.
The NRC staff has completed its review of the submitted evaluation and concludes that the structural integrity of the reactor pressure vessel at BFN1, is maintained. Additionally, the staff concludes that TVA has satisfied the requirements of 10 CFR 50.55a(b)(2)(xliii), 10 CFR 50.60, and 10 CFR Part 50, Appendix G. The staffs review is enclosed.
Sincerely, David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-259
Enclosure:
As stated cc: Listserv DAVID WRONA Digitally signed by DAVID WRONA Date: 2024.12.19 11:05:55 -05'00'
Enclosure STAFF EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO EVALUATION OF BROWNS FERRY NUCLEAR PLANT, UNIT 1 PRESSURE-TEMPERATURE EXCURSION FOLLOWING APRIL 2024 SCRAM TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-259
1.0 INTRODUCTION
By letter dated June 26, 2024 (Agencywide Documents Access and Management System Accession No. ML24179A028), the Tennessee Valley Authority (the licensee) submitted an evaluation of effects of out-of-limits condition that occurred following a reactor scram at Browns Ferry Nuclear Plant, Unit 1 (BFN1) to the U.S. Nuclear Regulatory Commission (Commission, NRC) for review and approval.
2.0 REGULATORY EVALUATION
Paragraph 50.55a(b)(2)(xliii) of Title 10 of the Code of Federal Regulations (prior to October 1, 2024) stated that Licensees shall submit for NRC review and approval the following analyses...(A) The analytical evaluation determining the effects of an out-of-limit condition on the structural integrity of the Reactor Coolant System [RCS], as described in IWB-3720(a).
The regulations in 10 CFR 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, require that all operating light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in 10 CFR Part 50, Appendix G, Fracture Toughness Requirements.
The regulations in 10 CFR Part 50, Appendix G, require: (1) sufficient fracture toughness for reactor pressure vessel (RPV) ferritic materials to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests; (2) pressure-temperature (P-T) limits that satisfy the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Appendix G, Fracture Toughness Criteria for Protection Against Failure, and the minimum temperature requirements during normal heatup, cooldown, and pressure test operations; and (3) applicable surveillance data from RPV material surveillance programs developed in accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, be incorporated into the calculations of P-T limits.
3.0 STAFF EVALUATION 3.1 Out-of-Limits Event In the submittal, the licensee stated that on April 24, 2024, during normal operation, BFN1 experienced a failed bushing in the main transformer, which tripped the turbine and subsequently initiated a reactor scram. Following the scram, BFN1 experienced a loss of feedwater pumps 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 RPV to maintain coolant level. The HPCI and RCIC systems only injected for a few minutes, but the CRD system continued to inject approximately 80 gallons per minute (gpm) directly to the bottom head of the RPV for several hours. This, in turn, resulted in temperatures of the RPV bottom head decreasing significantly with the remainder of the RPV which was still at nominal pressure. Later in the shutdown evolution, temperatures in the bottom head increased rapidly, exceeding specified heatup rate limits for the bottom head.
In the process of completing its standard RPV P-T surveillance monitoring, the licensee noted that RPV bottom head temperatures may have exceeded the Curve-1 P-T limits, and cooldown/heatup rates may have exceeded the 100 degrees Fahrenheit per hour (°F/hr) limit of the P-T limit curves in the BFN1 Technical Specification (TS) 3.4.9, RCS Pressure and Temperature (P/T) Limits.
3.2 Background
The BFN1 P-T limits curves are developed for 25 and 38 effective full-power years (EFPY) and are in TS 3.4.9. The P-T limit curves consist of the pressure test curve (Curve A or Curve-1),
heatup/cooldown conditions for core not critical curve (Curve B or Curve-2), and core critical operation curve (Curve C or Curve-3).
The P-T curves restrict the operation of the RPV to protect the RPV from brittle fracture. These curves are developed based on the American Society of Mechanical Engineers Boiler &
Pressure Vessel Code (ASME Code),Section XI, Appendix G and regulations in 10 CFR Part 50, Appendix G. These curves are developed based on linear elastic fracture mechanics to ensure that the BFN1 RPV is appropriately operated to protect its structural integrity. The flaw evaluation methodology is based on an assumed flaw of one-quarter of the RPV wall thickness (t), i.e., 1/4t flaw.
The ASME Code,Section XI, Appendix G specifies that the applied stress intensity factor (KI) must be less than the static fracture toughness (KIC) which is a material property of RPV shell materials. The available RPV material fracture toughness is defined by the initial RTNDT, a test-based property defined by NB-2300, Fracture Toughness Requirements for Material, of the ASME Code,Section III, Rules for Construction of Nuclear Facility Components. The RPV material fracture toughness can be reduced by neutron irradiation, which is predicted in the adjusted reference temperature (ART) for the RPV material. The neutron irradiation defines a value for KIC that decreases with increasing neutron irradiation. In addition, the KIC value of a material is affected by the operating temperature of a material, such that increasing temperature increases the value of KIC for a given value of RTNDT (or ART). The applied KI is affected both by the pressure in the RPV and by the thermal ramp rate.
The licensee stated that the RPV should be sufficiently warmed before pressurization for startup, and, during a shutdown, the pressure should be reduced before the vessel is allowed to cool. The specific parameters of these heatup and cooldown cycles are defined by P-T 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 P-T limit curves are defined by two applicability criteria. One is a time-based criterion for plant age (generally defined in EFPY) that reflects the accumulated embrittlement and shift in the KIC value for the RPV shell material. The other criterion is the temperature ramp rate, which provides a maximum thermal stress effect on the RPV shell material. Typically, the temperature ramp rate for the components in the RPV are limited to 100 °F/hr for heatup or cooldown, as shown in the P-T curves in the BFN1 TSs.
3.3 Licensees Analysis The licensee performed two evaluations. The first evaluation is a fracture mechanics analysis to demonstrate that the RPV (specifically, the bottom head) complies with 10 CFR Part 50, Appendix G via the requirements of the ASME Code,Section XI, Appendix G during the high-pressure, low-temperature state observed several hours after the initial scram event. The second evaluation is the development of P-T limit curves based on the out-of-limit heatup/cooldown rates to demonstrate that the bottom head P-T curve during the out-of-limits excursion is within the existing P-T limit curves.
3.3.1 Fracture Mechanics Analysis The licensee stated that the Code of Record for the construction of BFN1 is the 1965 Edition of the ASME Code. The lower head hemisphere is taken to be SA302 Grade B (Mn-1/2Mo) or SA508 Class 2 (C-1/2Mo-Cr-Ni). The licensee further stated that both materials have identical material properties for the thermal analysis (e.g., elastic modulus, mean coefficient of thermal expansion, diffusivity, specific heat, density, and Poissons ratio) based on the 1965 Edition of the ASME Code. The skirt attached to the bottom head is assumed to be SA106 Grade C, which, according to the licensee, is acceptable as the skirt is not the subject of this evaluation and is only modeled in the finite element analysis for completeness. The licensee obtained thermal diffusivity and specific heat from the 2017 Edition of the ASME Code,Section II, Materials, Part D, Properties. The inside diameter of the bottom head is 251 inches to the base metal. Cladding, which is not modeled, is 3/16 inches thick. The outside surface was assumed to be insulated with an ambient containment temperature of 100 °F. The NRC staff determined that not modeling the cladding is acceptable because it is more conservative in terms of thermal stresses than if cladding were to have been modeled in the stress analysis.
The licensee modeled two transients to calculate thermal stresses in the bottom head.
Transient-01 is modeled to mimic the as-measured temperature for the bottom head, specifically temperature at the sensor 56-30. The outside diameter temperatures were interactively checked and then compared against the BFN1 data.
Transient-02 is modeled for region C for the loss of feedwater pumps event, in which HPCI is defined to inject a total of three consecutive times.
The licensee performed its fracture mechanics analysis based on following steps:
Step 1: The licensee determined the fluid temperature, T, vs. time based on the actual thermal transient condition. The licensee used bottom thermocouple measurements to benchmark the calculated through-wall temperatures of the bottom head metal. The licensee calculated the through-wall temperatures in the bottom head at each time in the transient to determine the through-wall stresses and temperatures using the finite element analysis. The licensee stated that its finite element codes have been independently benchmarked against the ASME Code,Section XI, Appendix G method.
Step 2: The licensee postulated flaws at three locations (or paths) on the bottom head shell to ensure that the limiting condition in the bottom head has been evaluated. The licensee selected Path 1 at the juncture of the cylindrical shell and the spherical shell.
Path 2 is at the side of the spherical shell. Path 3 is at the bottom of the bottom head.
The licensee assumed a 1/4t depth axial crack model for Paths 1 and 2, and a 1/4t depth circumferential crack model for Path 3. For an assumed flaw at the limiting locations (paths), the licensee extracted the thermal stresses at the crack tip from the finite element analysis results to calculate the applied thermal stress intensity, KIT.
Step 3: The stresses in these three paths were then used in the fracture mechanics analysis. For each of these paths and at each point in time during the transient, the licensee calculated the static initiation fracture toughness, KIC, at the 1/4t depth crack tip temperatures based on the KIC equation specified in the ASME Code,Section XI, Appendix G.
Step 4: The licensee calculated applied stress intensity factor due to pressure (KIP). 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 (a is the crack depth and L is the crack length) and oriented normal to the largest primary membrane stress in the vessel head. The KIP is combined with KIT including the required structural factor to derive the total applied KI. Finally, the value of KI is compared to the value of KIC to determine whether the requirement of the ASME Code,Section XI, Appendix G is satisfied.
The NRC staff noted that for the Path 3 location, the licensee included an additional stress concentration factor (SCF) of 2.0 in the membrane (pressure) stress equation to account for stress concentrations at the holes for the CRD nozzles at the bottom of the bottom head. The licensee stated that 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.
However, page 2-17 of the NRC-approved topical report, BWROG-TP-11-022-A, Revision 1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, August 2013 (ML13277A557), recommends an SCF of 3.0 for the stress state around a hole, which is more conservative than the SCF of 2.0. The NRC staff performed an independent calculation applying the SCF of 3.0 on the membrane stress and determined that using the SCF of 3.0, the applied KI is still less than KIC and the requirement of the ASME Code,Section XI, Appendix G is satisfied.
The NRC staff verified that, as shown in Figures 5 through 8 in the Enclosure to the submittal, the licensee calculated applied crack tip stress intensity factors for pressure and thermal loading with appropriate structural factors, and they were all shown to be less than the allowable KIC values. The NRC staff determined that the applied stress intensity factor under the out-of-limits condition satisfies the ASME Code,Section XI, Appendix G requirements at all three limiting locations of the bottom head.
The NRC staff determined that although the higher ramp rate increases the thermal stress contribution to the applied stress intensity factor, the bottom head material retains sufficient margin in its KIC value as not to be negatively affected by the increased ramp rate. Therefore, the NRC staff finds that the out-of-limits condition does not significantly affect the structural integrity of the RPV.
3.3.2 Out-of-Limits P-T Curve The licensee developed an out-of-limits P-T curve for the bottom head using heatup or cooldown rates up to 200 °F/hr to determine whether the out-of-limits condition exceeds the existing P-T limits in the BFN1 TSs. The licensee developed this P-T curve based on the NRC-approved topical report, BWROG-TP-11-022-A, Revision 1. The licensee stated that the key criterion is that the ASME Code,Section XI, Appendix K, Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels, 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 RPV wall thickness and length equal to1.5 times the RPV wall thickness (i.e., aspect ratio a/L=1/6).
The licensee analyzed the extended beltline regions in the BFN1 RPV core mid-plane with fluence greater than 1 x 1017 neutrons per square centimeter (n/cm2) (E > 1.0 megaelectronvolt) for 50 EFPY (assumed EFPY for 80 years of plant operation). The licensee stated that no embrittlement shift from neutron irradiation is required for the bottom head material because it is located well below the extended beltline region. As such, the licensee used the initial RTNDT of 56 °F for the bottom head as the equivalent ART in developing the P-T curve. The NRC staff noted that the bottom head is in a RPV region where the neutron irradiation level is below 1 x 1017 n/cm2and finds it acceptable that irradiation embrittlement shift was not considered. The staff noted that irradiation embrittlement does not affect the RTNDT of the bottom head. The NRC staff finds that the unirradiated RTNDT of 56 °F can be used as the ART in the development of the out-of-limits P-T curve for the bottom head.
The licensee stated that the thermal contribution to stress intensity factor, KIT, for 100 °F/hr is 8.85 kilopounds per square inch square root inches (ksiinch), while the value for 200 °F/hr is 17.7 ksiinch. The NRC staff verified the calculation of KIT for the ramp rate of 200 °F/hr is 17.7 ksiinch and finds it acceptable for use in this assessment.
Aa shown in Figure 10 of the submittal, the out-of-limits P-T curve is positioned lower and to the right of the existing 100 °F/hr Curve 1 in the BFN1 TS 3.4.9. This shows that the out-of-limits P-T curve is more limiting than the 100 °F/hr Curve 1 in the TSs. The licensee stated that the new curve also provides margin to the existing Curve 2 (Curve B) for Core Not Critical operation in the TSs. The licensee further stated that all data associated with the ramp rate event are positioning to the lower right of the 200 °F/hr P-T curve, providing assurance of protection from brittle fracture.
The NRC staff noted that the ASME Code,Section XI, Nonmandatory Appendix E, Analytical Evaluation of Unanticipated Operating Events, permits an alternative method to analyze the out-of-limit condition on the structural integrity of the reactor vessel beltline region. Appendix E permits the use of 1.4 for the structural factor for the pressure stress and thermal stress, whereas ASME Code,Section XI, Appendix G specifies the structural factor of 2.0 for the pressure stress and 1.0 for the thermal stress. The NRC staff developed a P-T curve based on the Appendix E method and used the 200 °F/hr ramp rate. The NRC staff determined that the bottom head P-T curve developed by the licensee using the provisions of the ASME Code,Section XI, Appendix G is more conservative than the P-T curve developed by the NRC staff using the provisions of ASME Code,Section XI, Appendix E. Therefore, the licensees out-of-limits P-T curve for the bottom head as shown in Figure 10 of the submittal is acceptable.
4.0 CONCLUSION
Based on NRC staffs review of the licensees evaluation, the NRC staff has determined that (a) the calculated stresses at the bottom head are based on the temperature recordings from various sensors located on the RPV shell during the out-of-limits event, (b) the licensees fracture mechanics analysis demonstrated that the applied stress intensity factors are still within the material critical stress intensity factor (KIC), and (c) the bottom head P-T curve using the out-of-limit heatup/cooldown rate is within the existing P-T curve in the BFN1 TSs. The NRC staff has determined that the structural integrity of the RPV at BFN1 is maintained. The NRC staff concludes that the licensee has satisfied 10 CFR 50.55a(b)(2)(xliii), 10 CFR 50.60, and 10 CFR Part 50, Appendix G.
Principal Contributors: C. Moyer, NRR J. Tsao, NRR Date: December 19, 2024
ML24311A121 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DNRL/NVIB/BC NRR/DORL/LPLII-2/BC NAME KGreen ABaxter ABuford (OYee for)
DWrona DATE 11/06/24 11/15/24 10/11/24 12/19/24