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Introduction of ART Graphite R&D to include oxidation activities, oxidation resistant graphite, model development, ASME component failure, ASME code development (design rules), ceramic composites, AGC update, molten salt intrusion, split-disk studies, wear testing, and concluding remarks.
ASME Irradiation Model: How to deal with irradiation data in ASME code rules. Includes discussion on challenges to nuclear graphite, such as lack of sufficient irradiation data to qualify graphite for nuclear application for all grades and temperatures, variety of grades, and time/room in available MTRs to get all the required data. Discussion of Irradiation behavior, Leveraging data generated by GIF countries the past 20 years, Dimensional Change Theory, behavior, analysis, turnaround, and data. Future discussions include Arrhenius approach to predict turnaround behavior for the ASME MDS requirements for all nuclear graphite.
The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.
This report presents an evolutionary cyclic plasticity based transformative approach for fatigue evaluation of safety critical components. This approach is generic which can be used not only for the present generation light water reactor components (which is our focus) but also for advanced reactor components. The approach is based on the fundamental concept of time-evolution of progressive fatigue damage rather than the end-of-life data based conventional S~N curve approaches. Series of fatigue tests on 316 stainless steel uniaxial specimens were conducted under strain-controlled and stress-controlled cyclic loading in two different environment such as in-air at 300 °C and PWR coolant water condition at 300 °C. Material models are developed to capture the uniaxial test data and are subsequently programmed into commercially available ABAQUS code to transform the uniaxial material behavior to multiaxial loading domain. The developed evolutionary cyclic-plasticity approach is verified by both analytical and 3D-FE modeling of uniaxial fatigue test specimens. Results show that the developed material models can capture the time-dependence of material hardening and softening with great accuracy. In addition, results show that the material models not only can capture the material behavior under constant amplitude loading but also can capture the load-sequence effect under variable and random amplitude loading. The ANL developed approach is one of its kind, which shows first in the fatigue research community that fatigue evaluation of a component (in our case verified for laboratory scale fatigue test specimens with great accuracy) can be performed for its entire fatigue life, including the hardening and softening behavior, without using a conventional S~N curve based approach. Additionally results are presented related to fatigue evaluation of real reactor components such as of a pressurized water reactor (PWR) surge line (SL), under upset transient conditions. SLs of PWR experience complex thermal stratification due to the mixing of differential temperature flow from hot leg and pressurizer. A detailed parametric study of thermal stratification in SL is performed through both standalone computational fluid dynamics (CFD) analysis and fluid-structure-interaction based coupled CFD and heat transfer analysis. Results show that the mass flow rate in SL substantially affects the extent of its thermal stratification. The resulting thermal boundaries conditioned are used as inputs for thermal-mechanical stress analysis and component level fatigue evaluation. Fatigue evaluation of SL pipe under example loading conditions was performed using the above-mentioned evolutionary-plasticity based approach and ASME code (for in-air life) and NUREG-6909 (for PWR environment life) based approaches. Results show that the evolutionary plasticity based approach estimate substantially different life compared to the conventional ASME based approach. To show the entirety of the proposed evolutionary fatigue-modeling framework, this report includes the results presented in semi-annual March report (ANL/LWRS-17/01). Sections 2-5 of March report are directly included as Sections 2-5 in this report, without any significant changes. However, Sections 6-9 of this report present the work performed during March-September of 2017. Whereas, Sections 1 and 10 present the organization of this report and the summary and future work, respectively. This report (ANL/LWRS-17/03 Rev. 1) is a slightly revised version of the earlier published report (ANL/LWRS-17/03). In this revised version, some of the results and associated discussion have been corrected/updated.
The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, the secondary system in the 18-inch Test Vessel #6, and its sodium-to-air heat exchanger will occupy the location for 18-inch Test Vessel #5 (Test Vessel #5 will not be installed), Figure 2. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.
The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (like a redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed, most primary component have been fabricated, primary components are being assembled in preparation for initial testing, and procurement is continuing with the secondary system components. Since the FY19 THETA status report, The THETA primary system components are at Argonne National Laboratory (ANL) and have been dry assembled. The secondary sodium system for heat rejection from intermediate heat exchanger has all been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger has been procured.
The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed. Since the FY19 THETA status report, the THETA primary system components have been assembled and tested using de-ionized water. Some results of this testing at select Richardson and Reynolds numbers are included herein. The custom THETA sodium flowmeter (permanent magnet submersible sodium flowmeter) has been built and calibrated using sodium in an auxiliary facility against a NIST traceable vortex shedder type flowmeter. The permanent magnet flowmeter showed excellent performance, following the theoretical equation for measured voltage quite well, with an error < 3.5% at flow rates ranging from 3-24 GPM at temperatures up to 400 °C. The secondary sodium system for heat rejection from intermediate heat exchanger has been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger have been procured.
Environmental fatigue modeling is a complex problem due to multiple failure modes and their intermixing. The failure modes are function of various underlying causes in addition to the corrosive effect of reactor coolant environment. Some of the major causes are time-dependence of material associated with cyclic loading, load sequence effect associated with random/variable amplitude loading, effect of strain amplitude and rates, effect of varying temperature (along both temporal and spatial directions) and the effect of mean strain and stress. The nonlinear intermixing of failure modes associated with above mentioned causing parameters makes the environmental fatigue modeling is a challenging task. Because of this challenge, fatigue is traditionally being modeled based on experimental data. However, test based empirical approach often requires hundreds of fatigue tests to model the above-mentioned intermixing failure causes even for a single material system. The problem is further exaggerated for reactor component made from multi-material systems such as made from both carbon and stainless-steel base metals and their similar and dissimilar metal welds. With the difficulty of conducting hundreds of fatigue tests to capture the above-mentioned intermixing failure causes, fatigue modeling approaches often depends on empirical models based on limited available test data such as available through ASME code and NUREG 6909. However, these limited test-data-based models may not be enough to accurately predict the life of reactor components. Accurate prediction of life of reactor component would become a necessity, particularly when the license of the reactors to be extended for long-term-operation (LTO) that is for well beyond its original design life of 40 years. The requirement of extending the license of reactor under LTO requires hundreds of fatigue tests to be conducted to understand the mechanism associated with the above-mentioned interdependent failure causes. However, conducting large number of fatigue tests is not a feasibility due to the cost involved. To address this issues Argonne National Laboratory (ANL) with the sponsorship of DOE Light Water Reactor Sustainability (LWRS) program trying to develop a hybrid predictive modeling approach. This is based on limited experiment-data, Artificial-intelligence (AI) – Machine-Learning (ML) - Deep-Learning (DL) based techniques and Multiphysics-computational-mechanics based modeling tools. The hybrid approach not-only can improve the accuracy of the existing stress analysis and fatigue modeling approach but also can reduce the over-dependency on test-based approach. Towards this goal following are some of the major contributions based on ANL’s FY-20 environmental fatigue modeling activities: 1) A cyclic plasticity material model database for 82/182 dissimilar metal weld, which can be readily shared with US nuclear industry and regulatory agency on request. 2) A well validated analytical modeling methodology to perform cycle-by-cycle stress prediction under both constant amplitude fatigue loading and variable amplitude fatigue loading (with load-sequence effect). 3) An AI/ML/DL based methodology to predict unmeasurable cyclic strain based on other available sensor signals. This type of approach can be used for estimating strain in real reactor components from other sensor readings. 4) An AI/ML based approach to improve the US capability on environmental fatigue testing. This is by improving ANL’s existing environmental fatigue testing capacity to conduct ASME required strain-controlled tests (by controlling strain amplitudes and its rate), while not measuring the strain (due to the difficulty of placing an extensometer in a narrow autoclave in a PWR-water-test system). 5) A simulation and experiment based probabilistic modeling methodology for time-series fatigue state and life estimation of reactor metal such as dissimilar metal weld.
The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Futhermore, materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. As a result, this paper gives a summary of important results from the PIRT panel discussions and report.
Material property changes of fine- and medium-grain nuclear graphite grades were measured after subjection to uniform oxidation. The cores of high temperature reactors are composed of large nuclear graphite block components. Here, these large core components are designed to provide neutron moderation and reflection, create a large thermal sink to assist in operational control, and form the solid core structure containing the nuclear fuel, coolant channels, and the safety critical channels for control rod insertion. Oxidation is a principle degradation mechanism affecting all aspects of the nuclear graphite component functions. This study addresses the underlying physical property changes of nuclear-graphite components for oxidized mass loss ranges beyond the current recommended ASME code rule limits to ensure structural integrity within the graphite components (a maximum mass loss = 10%).
To understand high temperature creep-fatigue interaction of the Alloy 709, strain-controlled low-cycle fatigue (LCF) tests were performed at strain ranges varying from 0.3% to 1.2% with fully reversible cycle of triangular waveform at 750 °C. In addition, different hold times of 60, 600, 1800 and 3600 s were introduced at the maximum tensile strain to investigate the effect of creep damage on the fatigue-life at strain range of 1% at 750 °C. The creep-fatigue life and the number of cycles to macro-crack initiation and failure are found to decrease with increasing hold time indicating higher crack initiation and growth rates. Creep-fatigue life is evaluated by a linear summation of fractions of cyclic and creep damages according to ASME code. The fractographs of the samples deformed at 1% strain range indicated that fatigue might have been the dominant mode of deformation whereas, for the samples deformed at the same strain range with different hold times, both fatigue and creep have contributed to the overall deformation and fracture of the alloy.
We report at present, the available Ramberg–Osgood (R–O) parameters for different metals (e.g., in ASME code and other literatures) are static (generally based on a tensile curve). These static R–O parameters cannot accurately model the cyclic plasticity behavior. This work presents the cyclic R–O material hardening parameters for 316 stainless steel similar metal welds. The parameters were estimated under various conditions (in-air at room temperature, 300 °C in-air, and in-air at primary water conditions for a pressurized water reactor (PWR)). It is anticipated that the reported results would be useful for computational mechanics-based shakedown analysis and fatigue life estimation of PWR components.
Abstract Printed circuit heat exchangers (PCHEs) have high thermal efficiency because of the numerous minuscule channels. These minuscule channels result in a high thermal exchange area per unit volume, making PCHE a top contender for an intermediate heat exchanger (IHX) in high-temperature reactors. Thousands of minuscule channels make finite element analysis of the PCHE computationally infeasible. A two-dimensional analysis is usually performed for the PCHE core, which cannot simulate the local channel level responses reasonably because of the absence of global constraint influence. At present, there is no analysis technique available in the ASME Code or literature that is computationally efficient and suitable for engineers to estimate PCHE local responses. A novel but practical two-step analysis framework is proposed for performing PCHE analysis. In the first step, the channeled core is replaced by orthotropic solids with similar stiffness to simulate the global thermomechanical elastic responses of the PCHE. In the second step, local submodel analysis with detailed channel geometry and loading is performed using the elastic-perfectly plastic (EPP) material model. The proposed two-step analysis technique provides a unique capability to estimate the channel corner responses to be used for PCHE performance assessment. This study first developed a methodology for calculating the elastic orthotropic properties of the PCHE core. Next, the two-step analysis is performed for a realistic size PCHE core, and different issues observed in the results are scrutinized and resolved. Finally, a practical finite element analysis framework for PCHEs in high-temperature nuclear service is recommended.
High temperature nuclear reactors operating in the creep regime are designed to withstand numerous cyclic events. Current ASME code rules provide two basic paths for evaluating creep fatigue and ratcheting under these conditions; one based on full inelastic analysis intended to provide a representative stress and strain history and the other based on elastic material models with adjustments of varying complexity to account for inelastic stress and strain redistribution. More recent developments have used elastic-perfectly plastic analysis to bound the effects of cyclic service. However, these methods still rely on the separate evaluation of fatigue and creep damage utilizing a damage interaction diagram. There is a procedure under current development that uses creep-fatigue data from key feature test articles directly without the use of the damage interaction diagram. However, it requires a reasonable representation of the strain range in a structure as an input. This work develops a simplified procedure based on elastic perfectly-plasticity analysis that can be used to represent the strain range in a structure in the steady state under cyclic loading conditions.
The LBNF Absorber consists of thirteen 6061-T6 aluminum core blocks. The core blocks are water cooled with de-ionized (DI) water which becomes radioactive during beam operations. The cooling water flows through gun-drilled channels in the core blocks. A weld quality optimization was performed to produce National Aeronautical Standard 1514 Class I quality welds on the aluminum core blocks. This was not successful in all cases. An existing Gas Tungsten Arc Welding Procedure Specification was fine tuned to minimize, in most cases, and eliminate detect-able tungsten inclusions in the welds. All the weld coupons, however passed welding inspection as per the piping code: ASME B31.3 Normal Fluid Service. Tungsten electrode diameter, type, and manufacturer were varied. Some of the samples were pre-heated and others were not. It was observed that larger diameter electrodes, 5/32 in., with pre-heated joints resulted in welds with the least number of tungsten inclusions. It is hypothesized that thinner electrodes breakdown easily and get lodged into the weld pool during the welding process. This breakdown is further enhanced by the large temperature differential between the un-preheated sample and the hot electrode.
Creep-fatigue deformation is an important consideration for a thermal receiver of a gas phase (GP) Concentrating Solar Power (CSP) system due to the constant static stress or pressure, diurnal cycling, and elevated service temperatures required for efficient operation. An accurate description of the creep-fatigue behavior, not available for five of the six candidate materials, is important for assessment of preliminary designs. This project will provide a detailed analysis of the creep-fatigue behavior and damage accumulation of a candidate structural material for a CSP solar thermal receiver to address a critical knowledge barrier for receiver design in the GP pathway concept identified in the CSP Gen3 Demonstration Roadmap. This effort includes the development of rules for the design of solar receiver components against high temperature creep-fatigue and ratcheting failure modes. The ASME Code rules for high temperature nuclear components will form the basis of the method but adjustments will be made to reflect the generally shorter, diurnal operating cycles of thermal receivers and the relative consequences of failure, comparing nuclear to solar components. This is report covers the second quarter's work, which addressed outstanding questions on alloy selection for Quarter 1, and began laying the ground work for mechanical testing.
This report provides a summary on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18 (M2NT-18AN050502061), under the Work Package NT-18AN05050206, “G91 Code Extension Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors. The thermal aging study of G91 involves three types of thermally-aged specimens: (1) specimens machined from two heats of G91, G91-H1 and H30176 that have been tested for the Advanced Materials Development Programs in the past years. These specimens were aged at 550, 600, and 650°C for times up to ~64,000 h; (2) specimens fabricated from the heads of the archived crept specimens (13 in total) of six heats of G91, H5349, H30394, H30383, H10148, H30176. These specimens were aged at 427 - 538°C and for times up to 132,647 h; (3) specimens made from a tube removed from the Kingston coal-fired power plant after exposure for 155,000 h at 550°C. These three sets of specimens provide a comprehensive data set for understanding and predicting the effect of long-term thermal aging on microstructure and tensile properties of G91. Microstructural analysis of the thermally-aged G91 revealed that thermal aging causes significant microstructural changes: (1) the tempered martensite formed during tempering is unstable, and suffers significant recovery during aging, manifested by an increase in subgrain width, a decrease in subgrain length, and a decrease in subgrain aspect ratio (i.e. subgrains become more equiaxed) as the aging temperature and time increase. The subgrain recovery is facilitated by the reduction of dislocation density within subgrains; (2) M23C6 carbides in the as-received G91 are distributed along grain and subgrain boundaries and play an important role as pinning obstacles against subgrain coarsening. M23C6 carbides coarsen during aging, and the coarsening of M23C6 carbides reduces the pinning force on the boundaries; (3) MX carbonitrides in the as-received G91 are distributed uniformly within subgrains, providing the precipitation hardening effect. They coarsen with increasing aging temperature and time, but the coarsening rate is low; (3) a new phase, the Laves phase intermetallic forms during aging. The formation of the Laves phase removes the Mo solutes from the matrix, reducing the solid solution strengthening effect. The Laves phase particles grow rapidly once they are nucleated, and they precipitate primarily along boundaries. They may have a pinning effect on boundaries, but should provide little to the alloy strength. The evolution of each microstructural constituent during thermal aging can be well described by individual microstructural models. The estimated activation energy for MX coarsening was 175 kJ/mol, equivalent to the activation energy for boundary and core diffusion in a-Fe, while the estimated activation energy for subgrain coarsening was significantly lower, ~96 kJ/mol. Thermal aging of G91 results in the reduction in the yield stress and the ultimate tensile strength but has an insignificant effect on the uniform and total elongations. The reduction in tensile strength is more pronounced as the aging temperature and time increases. The in situ tensile tests with high-energy X-rays revealed that sub-boundary strengthening plays a dominant role, and the strengthening role of M23C6 particles may be accounted for by subgrian boundary hardening, and three strengthening mechanisms can be superimposed to describe the tensile strength of G91: subgrain boundary hardening, MX precipitation hardening, and Mo solid solution strengthening. A microstructure – strength model was establ
Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. The present semi-annual report covers the period between May 2018 and September 2018. In this report, we present the work performed to further improve the capability for structural integrity prediction. Part of this work focused on validating a computational fluid dynamics model for modeling thermal stratification. We also present finite element (FE) model based thermal-mechanical stress analysis of a pressurized water reactor (PWR) surge line under grid-load-following loading condition. In addition, we present work related to end-of-life prediction of a PWR surge line based on ASME code and NUREG-6909 deterministic approaches under design-basis and grid-load-following conditions. Our research in this report period also majorly involved with the development of a strategy for probabilistic fatigue life estimation. First, we applied a Weibull probabilistic modeling approach based on end-of-life fatigue data (traditional S~N data under in-air and PWR water conditions) to estimate the probabilistic life for a given strain amplitude. Second, we developed a data analytics based Markov-Chain-Monte-Carlo (MCMC) model for probabilistic modeling of time-series fatigue damage states and the associated probabilistic end-of-life of a component. Unlike the Weibull type approach, the MCMC type may not require a large number of fatigue tests for estimating the end-of-life for given loading and environmental conditions. We demonstrated the use of this framework through symmetric fatigue tests (stress- or strain-ontrolled tests either under in-air or PWR-water conditions) obtained from our earlier experiments. The MCMC model can use base time-series data either obtained from a model or fatigue experiment. Since experimental data are typically more accurate than model-calculated data, we used experiment-based time-series data to demonstrate the MCMC framework. However, when performing a complex experiment is not possible (due to time and costs involved), a model-based approach can be used to generate the time-series data. Similar model-based data are presented in our earlier report (e.g. for cyclic hardening damage states of a PWR surge line under constant amplitude and design-basis loading [1]). However, for the MCMC work in this report, we used experiment-based time-series damage states obtained from fatigue experiments under design-basis and grid-load-following loading conditions. The experimental loading inputs were based on thermal-mechanical FE model results of a typical PWR surge line. Then, we used the experimental time-series damage state data for forecasting the probabilistic time-series damage states and end-of-life of the PWR surge line. Based on this MCMC modeling strategy, and with the assumed loading, environment and thermal-mechanical boundary conditions, we have estimated that a 316SS-pure base metal PWR surge line would survive at least 159 years with zero failure probability.