Examples of TLAA Evaluations

Revision 0

These are examples of TLAAs that were in license renewal applications and have been accepted by the NRC. The plant specific names have been omitted, but specific references to industry standards and other publicly available information have been retained. These examples include the three approaches in 10 CFR 54.21(c)(1) to disposition the TLAAs.

10 CFR 54.21 (c)(1)(i) - Example (The analyses remain valid for the period of extended operation)

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REACTOR BUILDING CRANE CYCLIC LOADING ANALYSIS

TLAA Description:

The reactor building crane is common to both units at the plant and is within the scope of license renewal. It was designed to meet the fatigue requirements of the ASME NOG-1-2004 and Crane Manufacturer Association of America (CMAA) Specification 70 for a Class A, Standby or Infrequent Service Crane, as discussed in the plant's UFSAR Section 9.1.4.2.3, Reactor Building Crane. This evaluation of load cycles over the 40-year plant life is the basis of a safety determination and has been identified as a TLAA that requires evaluation for the period of extended operation.

TLAA Evaluation:

The evaluation of the reactor building crane cyclic load limit TLAA included (1) reviewing the existing 40-year design basis to determine the number of load cycles considered in the design of the crane, (2) developing a 60-year projection for load cycles for the crane, and (3) comparing the 60-year projected number of load cycles to the 40-year design load cycles.

The reactor building crane is designed in accordance with CMAA Specification 70. Referring to Table 2.8-1 of CMAA Specification 70 (2004), the reactor building crane is a Class A crane and can be considered a crane experiencing “irregular occasional use followed by long idle periods.” For this crane, the CMAA design considerations allow for between 20,000 and 100,000 load cycles. Therefore, 20,000 load cycles is a conservative limitation on load cycles for his crane. Load cycles that lift less than 50 percent of the crane design capacity of 125 tons (62.5 tons) result in minimal fatigue of the crane. Therefore, load cycles that lift 50 tons or more are evaluated.

Table 4.7.1-1 provides the 60-year projections for reactor building crane load cycles. The number of load cycles projected for 60 years of operation is 2,672 load cycles.

The 60-year projected number of load cycles is less than 20 percent of the allowable design value of 20,000 load cycles. Therefore, the reactor building crane load cycle fatigue analysis remains valid for 60 years of plant operation.

Table 4.7.1-1
Unit 1 and Unit 2 Reactor Building Crane Load Cycles
Heavy Load Description	Frequency	Number of years	Total Cycles
Construction Total:	200
Refueling Outage Cycles:
Reactor Vessel Head	2 / year	60	120
Drywell Head	2 / year	60	120
Reactor Vessel Steam Separator	2 / year	60	120
Reactor Cavity Shield Plugs (6)	12 / year	60	720
Dryer / Separator Shield Plugs (1)	2 / per year	60	120
Miscellaneous	4 / year	60	240
Completed ISFSI Cask Load Cycles (16 Casks)	96
ISFSI Cask Load Cycles (2015 – 2043)	30 / year	29	870
Completed Low Level Waste Cask Load Cycles (4 Casks)	24
Low Level Waste Cask Load Cycles – 7 Casks (2015 – 2043)	42
60-year Total Load Cycles	2,672
Design Load Cycles	20,000
Percent of Design Load Cycles after 60 years	13.30%

TLAA Disposition: 10 CFR 54.21(c)(1)(i) - The analysis remains valid for the period of extended operation.

10 CFR 54.21 (c)(1)(ii) - Example (The analyses have been projected to the end of the period of extended operation)

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MAIN STEAM LINE FLOW RESTRICTORS EROSION ANALYSIS

TLAA Description:

A main steam line flow restrictor is welded into each of the four main steam lines between the main steam relief valves and the inboard main steam isolation valve. The restrictor assemblies consist of a stainless steel venturi-type nozzle welded into the carbon steel main steam line piping. The restrictors are designed to limit steam flow to less than 200 percent prior to MSIV closure in the event of a main steam line break outside of primary containment to limit reactor coolant loss, maintain core cooling, and limit the release of radiological material to the environment within allowable regulatory limits.

The analysis of main steam line flow restrictor erosion is discussed in the Updated Final Safety Analysis Report (UFSAR) Section 5.4.4, summarized below. UFSAR Section 5.4.4 indicates that very slow erosion of the venturi occurs with time and such slight enlargement has no safety significance. Since the erosion evaluation was based on 40 years of operation, erosion of the main steam line flow restrictors has been identified as a TLAA that requires evaluation for the period of extended operation.

TLAA Evaluation:

The resistance of stainless steel to erosion has been established by turbine inspections at another BWR plant that revealed no noticeable effects from erosion on the stainless steel nozzle partitions at similar steam velocities. Calculations indicate that even with erosion rates as high as 0.004 inch per year, after 40 years of operation the increase in choked flow rate would be no more than 5 percent.

The main steam line break accident is discussed in UFSAR Section 15.6.4. The assumed integrated mass of coolant leaving the reactor through the main steam line break is 100,000 lb., of which 14,000 lb. is liquid and 86,000 lb. is steam. The original design bases analyses for the radiological consequences of the main steam line break, summarized in UFSAR Table 15.6-8, resulted in 0.0354 rem whole body dose at the exclusion area boundary in the first 2 hours, and 0.000015 rem whole body dose at the low population zone after 30 days; well below the 10 CFR 100 limit of 25.0 rem. Even if the choked flow rate is increased an additional 5 percent to account for the additional 20 years of service, for a total of 10 percent the increase in dose consequences is negligible relative to 10 CFR 100 limits. Therefore, the potential loss of material due to erosion has been projected to the end of the period of extended operation with acceptable results.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii) - The analysis has been projected to the end of the period of extended operation.

10 CFR 54.21 (c)(1)(iii) - Example (The effects of aging on the intended function(s) will be adequately managed for the period of extended operation)

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ENVIRONMENTAL QUALIFICATION (EQ) OF ELECTRIC COMPONENTS

TLAA Description:

Thermal, radiation, and cyclical aging analyses of plant electrical and I&C components, developed to meet 10 CFR 50.49 requirements, have been identified as TLAAs for the plant. The US NRC has established nuclear station environmental qualification (EQ) requirements in 10 CFR 50.49 and 10 CFR 50, Appendix A, Criterion 4. 10 CFR 50.49 specifically requires that an EQ program be established to demonstrate that certain electrical components located in harsh plant environments are qualified to perform their safety function in those harsh environments after the effects of in-service aging. Harsh environments are defined as those areas of the plant that could be subject to the harsh environmental effects of a loss-of-coolant accident (LOCA), high energy line break (HELB), or post-LOCA radiation. 10 CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of environmental qualification.

Environmental Qualification Program Background

The plant's EQ Program meets the requirements of 10 CFR 50.49 for the applicable electrical components important to safety. 10 CFR 50.49 defines the scope of components to be included, requires the preparation and maintenance of a list of in scope components, and requires the preparation and maintenance of a qualification file that includes component performance specifications, electrical characteristics and the environmental conditions to which the components could be subjected.

10 CFR 50.49 (e)(5) contains provisions for aging that require, in part, consideration of all significant types of aging degradation that can affect component functional capability. 10 CFR 50.49 (e)(5) also requires replacement or refurbishment of components not qualified for the current license term prior to the end of designated life unless additional life is established through ongoing qualification. 10 CFR 50.49(f) establishes four methods of demonstrating qualification for aging and accident conditions. 10 CFR 50.49(k) and (l) permit different qualification criteria to apply based on plant and component vintage. Supplemental EQ regulatory guidance for compliance with these different qualification criteria is provided in NUREG-0588, Revision 1, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," July 1981 (Reference 4.8.15), and Regulatory Guide 1.89, Revision 1, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants," June 1984 (Reference 4.8.16).

Compliance with 10 CFR 50.49 provides reasonable assurance that the component can perform its intended functions during accident conditions after experiencing the effects of in-service aging. The plant's EQ Program manages component thermal, radiation, and cyclical aging, as applicable, through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current license term are to be refurbished, replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation.

Aging evaluations for electrical components in the plant's EQ Program that specify a qualification of at least 40 years are TLAAs for license renewal because the criteria contained in 10 CFR 54.3 are met.

TLAA Evaluation:

The plant's EQ Program implements the requirements of 10 CFR 50.49, as further defined and clarified by NUREG-0588, Revision 1 and Regulatory Guide 1.89, Revision 1, and is viewed as an aging management program for license renewal under 10 CFR 54.21(c)(1)(iii). Reanalysis of an aging evaluation to extend the qualifications of components is performed on a routine basis as part of the plant's EQ Program. Important attributes for the reanalysis of an aging evaluation include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions (if acceptance criteria are not met). TLAA demonstration option (iii), which states that the effects of aging will be adequately managed for the period of extended operation, is chosen and the plant's EQ Program will manage the aging effects of the components associated with the environmental qualification TLAA.

NUREG-1800 states that the US NRC evaluated the EQ program (10 CFR 50.49) and determined that it is an acceptable aging management program to address environmental qualification according to 10 CFR 54.21(c)(1)(iii). The evaluation referred to in the [Standard Review Plan for License Renewal contains sections on “EQ Component Reanalysis Attributes, Evaluation, and Technical Basis” that is the basis of the description below.

Component Reanalysis Attributes

The reanalysis of an aging evaluation is normally performed to extend the qualification by reducing conservatism incorporated in the prior evaluation. Reanalysis of an aging evaluation to extend the qualification of a component is performed on a routine basis pursuant to 10 CFR 50.49(e) as part of the plant's EQ Program. While a component life-limiting condition may be due to thermal, radiation, or cyclical aging, the majority of component aging limits are based on thermal conditions. Conservatism may exist in aging evaluation parameters, such as the assumed ambient temperature of the component, unrealistically low activation energy, or in the application of a component (de-energized versus energized). The reanalysis of an aging evaluation is documented according to the plant's quality assurance program requirements, which require the verification of assumptions and conclusions. As previously noted, important attributes of a reanalysis include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions (if acceptance criteria are not met). These attributes are discussed below.

Analytical Methods

The plant's EQ Program uses the same analytical models in the reanalysis of an aging evaluation as those previously applied during the prior evaluation. The Arrhenius methodology is an acceptable thermal model for performing a thermal aging evaluation. The analytical method used for a radiation aging evaluation is to demonstrate qualification for the total integrated dose, which is the normal radiation dose for the projected installed life plus accident radiation dose. For license renewal, one acceptable method of establishing the 60-year normal radiation dose is to multiply the 40-year normal radiation dose by 1.5 (that is, 60 years/40 years). The result is added to the accident radiation dose to obtain the total integrated dose for the component. For cyclical aging, a similar approach may be used. Other models may be justified on a case-by-case basis.

Data Collection & Reduction Methods

The chief method used for a reanalysis per the plant's EQ Program is reduction of conservatism in the component service conditions used in the prior aging evaluation, including temperature, radiation, and cycles. Temperature data used in an aging evaluation is conservative and based on plant design temperatures or on actual plant temperature data. When used, plant temperature data can be obtained in several ways, including monitors used for technical specification compliance, other installed monitors, measurements made by plant operators during rounds, and temperature sensors on large motors. A representative number of temperature measurements are evaluated to establish the temperatures used in an aging evaluation. Plant temperature data may be used in an aging evaluation in different ways, such as: (a) directly applying the plant temperature data in the evaluation or (b) using the plant temperature data to demonstrate conservatism when using plant design temperatures for an evaluation. Any changes to material activation energy values as part of a reanalysis must be justified. Similar methods of reducing conservatism in the component service conditions used in prior aging evaluations can be used for radiation and cyclical aging.

Underlying Assumptions

The plant's EQ Program component aging evaluations contain sufficient conservatism to account for most environmental changes occurring due to plant modifications and events. When unexpected adverse conditions are identified during operational or maintenance activities that affect the normal operating environment of a qualified component, the affected EQ component is evaluated and appropriate corrective actions are taken, which may include changes to the qualification bases and conclusions.

Acceptance Criteria and Corrective Action

Under the plant's EQ Program, the reanalysis of an aging evaluation could extend the qualification of the component. If the qualification cannot be extended by reanalysis, the component is refurbished, replaced, or requalified prior to exceeding the period for which the current qualification remains valid. A reanalysis is to be performed in a timely manner such that sufficient time is available to refurbish, replace, or requalify the component if the reanalysis is unsuccessful.

The plant's EQ Program has been demonstrated to be capable of programmatically managing the qualified lives of the components within the scope of the program for license renewal. The continued implementation of the plant's EQ Program provides reasonable assurance that the aging effects will be managed and that EQ components will continue to perform their intended functions for the period of extended operation. A comparison of the plant's Environmental Qualification (EQ) of Electric Components (B.3.1.3) program to the corresponding program in NUREG-1801 is provided in Appendix B, Subsection B.3.1.3.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii) - The effects of aging on the intended function(s) will be managed by the plant's EQ Program for the period of extended operation.

Record of Revisions

Number	Date	Description of Changes
0	6/11/2024	Initial version

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

Code of Federal Regulations - The US federal government regulations. 10 CFR 54 is the regulation for renewal of nuclear power plant operating licenses.

American Society of Mechanical Engineers

Crane Manufacturers Association of America

Updated Final Safety Analysis Report - The Updated Final Safety Analysis Report is the version of the FSAR that has been revised during the operation of the plant and is the most recent version.

Boiling Water Reactor

Environmental Qualification - A process for ensuring that equipment will be capable of withstanding the ambient conditions that could exist when the specific function to be performed by the equipment is actually called upon to be performed under accident conditions.

Instrumentation and Control

United States of America

Loss of Coolant Accident

High Energy Line Break - A postulated event of ruptures of high energy piping.