Pressure-temperature Limits Research Articles

Development of pressure-temperature limit curves considering unified constraint for reactor pressure vessel

In this work, the pressure-temperature (P-T) limit curves considering unified constraint for a typical large-sized reactor pressure vessel (RPV) with axial surface cracks have been investigated and developed based on the KJ-Ad* two-parameter and the Master Curve (MC) method. The ASME fracture toughness KIc curve and the MC with and without considering unified constraint for irradiated RPV steel were used in the constructions of the P-T limit curves. The effects of crack sizes (constraint levels) and fracture toughness curves on the P-T limit curves have been analyzed for the RPV under heatup and cooldown conditions. The results further show that the P-T limit curves based on traditional ASME KIc curve has the highest degree of conservatism and the use of the MC method can significantly increase the safety margin of P-T limit curves. With increasing crack sizes (increasing constraint levels), the safety region in P-T limit curves reduces, which may lead to the decrease of operating flexibility for nuclear power plants. The accuracy of the P-T limit curves can be improved by considering unified constraint. For smaller size cracks with lower constraint in the RPV, the consideration of constraint can reduce conservatism of the P-T limit curves, and for larger size cracks with higher constraint, it can avoid the non-conservatism. It is recommended to obtain more accurate P-T limit curves based on the MC with considering unified constraint and finite element analysis for RPVs under various operating conditions.

Structural Integrity Assessment of a Nuclear Vessel through ASME and Master Curve Approaches Using Irradiation Embrittlement Predictions

Abstract The assessment of the structural integrity of nuclear vessels is based on a series of procedures developed in the 1970s and 1980s. On one hand, curves that, according to the American Society of Mechanical Engineers code, describe the relationship between steel toughness and temperature in the ductile-to-brittle transition region, based on the reference temperature concept RTNDT, were adopted in 1972. On the other hand, the material embrittlement derived from the exposure of steel to neutron irradiation is determined through the model included in “Regulatory Guide 1.99 Rev. 2,” published in 1988. Since then, there have been enormous advances in this field. For example, the Master Curve, based on the reference temperature T0, describes the relationship between toughness and temperature in the transition zone more realistically and with much more robust microstructural and mechanical foundations and uses the elastic-plastic fracture toughness KJc. Moreover, improved models have been developed to estimate the embrittlement of steel subjected to neutron irradiation, such as ASTM E900, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials. This study is aimed at comparing the results obtained using traditional procedures to the improved alternatives developed later. For this purpose, the behavior of the steel of a nuclear vessel that is currently under construction has been experimentally characterized through RTNDT and T0 parameters. In addition, the material embrittlement has been quantified using “Regulatory Guide 1.99 Rev. 2” and ASTM E900. These experimental results have been transferred to the assessment of the structural integrity of the vessel to determine the pressure-temperature limit curves and size of the maximum admissible defect as a function of the operation time of the plant. The results have allowed the implicit overconservatism present in the traditional procedures to be quantified.

Thermal response of high-aspect-ratio hydrogen cylinders undergoing fast-filling

Fast-filling can lead to excessive heating of a hydrogen cylinder. Gas cylinders with large length-to-diameter aspect ratios are prone to developing hot-spots that may cause the temperature of the structure locally to exceed specified limits. The hot-spots develop because convective heat transfer into the cylinder wall is not uniform along the length of the cylinder. Computational fluid dynamic simulations of fast-filling reveal that the flow within large aspect ratio cylinders is characterised by two distinct flow patterns. The region near to the inflow nozzle exhibits an axi-symmetric recirculating flow pattern that extends to approximately three cylinder diameters downstream from the inlet. Wall heat transfer in the recirculation region of the flow is driven strongly by the jet of turbulent fluid from the nozzle. The wall heat transfer in the recirculation region rapidly develops quasi-steady behaviour in which the Nusselt number is approximately proportional to the inlet jet Reynolds number. Downstream of the recirculation zone there is a region of axial flow, in which wall heat transfer occurs at a slower rate, driven by decaying turbulence transported from the recirculation zone. The reduced heat transfer in the end gas leads to a higher gas temperature but also to a lower structural temperature downstream of the recirculation zone. Analysis shows that wall heat transfer can be enhanced and peak gas temperatures reduced by dividing the hydrogen injection between separate nozzles. Directing the nozzles axially, ideally at intervals of around 4 cylinder diameters, increases heat transfer and prevents the formation of hot regions of stagnant fluid. This approach allows more hydrogen to be stored for a given pressure or gas temperature limit. However the analysis across various length-to-diameter ratios and nozzle configurations reveals substantially different trends for the gas temperatures appearing in ISO safety specifications, and the temperature of the structural materials they seek to safeguard.

Development of pressure-temperature operation limits for a PWR vessel considering beltline shell and extended beltline nozzles

The pressure-temperature limits (P-T limits) of embrittled reactor pressure vessels (RPVs) should be determined for normal heat-up and cool-down operations based on fracture mechanics to ensure the structural integrity. In general, the P-T limits are mainly determined by the fracture toughness of beltline region material which suffers the highest level of neutron embrittlement. However, some other components such as nozzles which suffer insignificant neutron embrittlement may restrict the P-T limits as a result of higher stress concentration from their structural discontinuities. Therefore, beltline region material with the highest reference temperature as well as nozzles that have structural discontinuities need to be considered when developing the P-T limits of RPV. In this paper, the P-T limit calculations for a Taiwan domestic pressurized water reactor (PWR) vessel were performed considering both beltline shell and extended beltline nozzles according to the nonmandatory Appendix G to ASME Code Section XI. The three-dimensional finite element analyses were conducted to obtain the pressure and thermal stress distributions of extended beltline nozzles for P-T limits calculation. It is found that the beltline region still dominates the cool-down P-T limit, but the extended beltline region will partially control the heat-up limit. Further, the reference temperature ranges of extended beltline nozzles that may influence the P-T limits of RPV were also investigated. Present study is helpful for safe operation and also provides a reference for regulation of PWR plants in Taiwan.

Effect of postulated crack location on the pressure-temperature limit curve of reactor pressure vessel

Abstract In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the KIa fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the KIC fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan.

Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

Abstract The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ( RT NDT ). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

Effects of fracture toughness curves of ASME Section XI–Appendix G on a reactor pressure vessel under pressure–temperature limit operation

According to the Code Case N-640 issued in 1999, the fracture toughness requirement of reactor pressure vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present pressure–temperature limit operation curves of normal reactor startup (heat-up) and shut-down (cool-down) for the reactor pressure vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic reactor pressure vessel under various pressure–temperature limit operations were analyzed. First, the pressure–temperature limit curves of the reactor pressure vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the reactor pressure vessel under such pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the pressure–temperature limit transients based on KIC curves, the reactor pressure vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed reactor pressure vessel.

LIMITS ON QUAOAR'S ATMOSPHERE

Here we present high cadence photometry taken by the Acquisition Camera on Gemini South, of a close passage by the $\sim540$ km radius Kuiper Belt Object, (50000) Quaoar, of a r'=20.2 background star. Observations before and after the event show that the apparent impact parameter of the event was $0.019\pm0.004$", corresponding to a close approach of $580\pm120$ km to the centre of Quaoar. No signatures of occultation by either Quaoar's limb or its potential atmosphere are detectable in the relative photometry of Quaoar and the target star, which were unresolved during closest approach. From this photometry we are able to put constraints on any potential atmosphere Quaoar might have. Using a Markov chain Monte Carlo and likelihood approach, we place pressure upper limits on sublimation supported, isothermal atmospheres of pure N$_2$, CO, and CH$_4$. For N$_2$ and CO, the upper limit surface pressures are 1 and 0.7 $\mu{bar}$ respectively. The surface temperature required for such low sublimation pressures is $\sim33$ K, much lower than Quaoar's mean temperature of $\sim44$ K measured by others. We conclude that Quaoar cannot have an isothermal N$_2$ or CO atmosphere. We cannot eliminate the possibility of a CH$_4$ atmosphere, but place upper surface pressure and mean temperature limits of $\sim138$ nbar and $\sim44$ K respectively.

Mechanical Extrusion Offers Advantages Over Ball Activation for Downhole Tools

Technology Update One of the most innovative recent advances in downhole drilling tool activation has been the introduction of mechanical extrusion technology, incorporating the use of rigid metal darts to improve the cycling and control of a variety of processes and devices. Increasingly, operators are recognizing mechanical extrusion by means of dart activation as a time-, cost-, and labor-saving alternative to traditional ball activation methods. For example, conventional cycling of drilling bypass valves has used a polymer extrusion process that relies on balls as the activation device. However, Churchill Drilling Tools has developed an effective and accurate mechanical extrusion method that uses a dart-based valve. The system can perform activations up to five times faster than polymer methods. The company’s MX (mechanical extrusion) system allows the use of rigid Smart Darts for multicycle control and exploits their robustness and resilience at high pressures and temperatures (up to 660°F) to deliver greater operating speed, reliability, and performance. The system’s unique feature is its two configurable shearing modes, which solves reliability and power delivery problems. It requires a radius as small as 0.0075 in. for power transmission, thereby making implementation extremely compact. With the ability to control and power three or more tools independently, users eliminate the conventional inner diameter (ID) and obstruction conflicts between different tools in a string. This has implications for mono-cycling ball and shear systems, which commonly obstruct the bore after use. These can be turned into multicycling, nonobstructive, and multi-tool systems. The system can also enhance multicycling ball-activated tools that are based on polymer extrusion by improving performance and delivering greater multitool compatibility and interoperability. The system allows the mechanical extrusion to be delivered in a range of settings to suit the specific cycle application. With a variety of latching, sealing, and flow path geometries, as well as the adjustable shear mode ratings, this element of the system provides a tangible change in string design and program implementation. For example, in bypass valve implementation, users can cover multiple flow path contingencies with a single valve in hole and choose from the range of darts to establish the optimal setting for the situation encountered. The system advances the capability of simple hydromechanical control and its power and reliability benefits into areas that previously needed more complex electronic systems. Being mechanical, it is not subject to the same high pressure and high temperature limits that affect electronics and electrical components. Smart Darts enable a multicycle circulating sub (circ sub) tool to deliver a reliable and versatile bypass on demand. Whatever the drilling application, whether planned or a contingency, the system enables a rapid and reliable switching to the optimal flow path configuration.

A Risk-Informed Methodology for ASME Section XI, Appendix G

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

Experimental synthesis and stability of tourmaline: a historical overview

Experiments with tourmaline-group minerals began with synthesis, and later, synthesis with the aim of defining pressure–temperature limits on the stability fields of end-member compositions. As a result, experiments have confirmed what has been evident from natural occurrences, that the common tourmaline-group minerals (solid solutions among schorl, dravite, olenite, foitite and, to a lesser extent, uvite) can be stable from the conditions of diagenesis through the granulite facies. However, the stability field of elbaite, the most common Li-rich species, is not known with any certainty, and most attempts at its synthesis have been unsuccessful. No tourmaline-group mineral has yet been synthesized at atmospheric pressures. A second wave of experimentation focused on the chemical attributes of environments that determine the composition and stability of tourmaline. Most notably, the stability (saturation) of tourmaline has been related to the boron content of melts and fluids, to aluminosity of melt and alkalinity of fluid, to the activities of Y -site cations, oxidation state, and activities of H 2 O and F in fluids and melts. As a result, the chemical stability of common tourmaline compositions can now be explained in terms of the activity or concentration products of the constituent chemical components for each structural site. Experimental studies of isotopic exchange between tourmaline and aqueous fluid have focused mostly on Δ 11 B tourmaline–vapor. There is one experimental study of ΔD tourmaline–vapor, but Δ 18 O tourmaline–vapor or melt has not yet been measured. The absence of an experimental calibration Δ 18 O tourmaline–vapor or tourmaline–melt is a major limitation on the utilization of tourmaline for geothermometric purposes.

The Feasibility of Using a Risk-informed Approach for Calculating Reactor Pressure Vessel Heatup and Cooldown Operating Curves

Abstract The current procedures for calculating pressure-temperature (P/T) limits for normal reactor startup and shutdown are defined by deterministic fracture mechanics methodology in the ASME Code, Appendix G (in both Section XI and Section III). The recent pressurized thermal shock (PTS) re-evaluation effort used a very thorough probabilistic fracture mechanics (PFM) evaluation to develop a technical basis to increase the PTS screening criteria. The feasibility of applying this same PFM methodology to evaluate normal startup and shutdown operation for both pressurized water reactor (PWR) and boiling water reactor (BWR) pressure vessels is described in this paper. The approach taken in this study was to define a new risk-informed margin term to be applied to the stress intensity factor for membrane tension (KIm) in Appendix G. The margin on KIm was determined by finding the value that results in a vessel failure frequency equal to 10−6 failures/reactor-year when the reactor operates up to the pressure-temperature limits calculated using the risk-informed margin. This simple approach was selected because it results in a minimum change to the current ASME Code procedure, is easy to apply when revising plant operating P/T limits, and allows for an increase in the allowable P/T limits by as much as would be provided by any alternate procedure. The results from this initial study indicate the margin on KIm in Appendix G can be reduced from 2 to 1.5 for PWR vessels for shutdown and potentially reduced from 2 to as low as 1 for startup. The results indicate the margin on KIm in Appendix G can be reduced from 2 to approximately 1 for BWR vessels for shutdown and potentially startup. Additional analyses for PWR and BWR vessels will be needed to develop a comprehensive risk-informed basis for any revisions to the ASME Code, Appendix G.

Determination of Pressure-Temperature Limit Curves Using Deterministic and Probabilistic Approaches

Reactor pressure vessel (RPV) is the most critical component in nuclear power plant. RPV is subjected to radiation embrittlement, which is characterized as neutron fluence-dependent reduction in fracture toughness of the material. Therefore, risk for potential failure of RPV increases as operating time and fluence level increase. To prevent the potential failure, it is requested for RPV to operate in accordance with pressure-temperature (P-T) limit curve during operation. However, it has been reported that P-T limit curve which is typically developed in accordance with the procedure in ASME code is too conservative. Therefore, in order to investigate the conservatism of current P-T limit curve and develop more realistic one, probabilistic approach based on the risk was utilized in this paper. The resulting P-T limit curve is very higher than that from deterministic approach, and can be used as alternative operation limit of the RPV, because probabilistic P-T limit curve seems to have enough safety margin for potential failure of RPV.

Effects of Major Parameters on the P-T Limit Curve Construction of Embrittled Reactor Vessel

To prevent the potential failure of the reactor pressure vessel (RPV), it is requested to operate RPV according to the pressure-temperature (P-T) limit curve during the heat-up and cool-down process. The procedure to make the P-T limit curve was suggested in the ASME Code but it has been known to be too conservative for some cases. In this paper, the conservatism of the ASME Code Sec. XI, App. G was investigated by performing a series of sensitivity analyses. The effects of six parameters such as crack depth, crack orientation, clad thickness, fracture toughness, cooling rate, and neutron fluence were analyzed. The results of P-T limit curves are compared to one another.

Pressure-temperature limit curve for reactor vessel evaluated by ASME code

A comparative assessment study for a generation of the pressure-temperature (P-T) limit curve of a reactor vessel is performed in accordance with ASME code. Using cooling or heating rate and vessel material properties, stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during reactor cool-down and heat-up. P-T limit curves are analyzed with respect to defect orientation, clad thickness, toughness curve, cooling or heating rate and neutron fluence. The resulting P-T curves are compared each other.

Alternative Method of RTNDT Determination for Some Reactor Vessel Weld Metals Validated by Fracture Toughness Data

Abstract The fracture toughness curves used for nuclear power plant operation pressure-temperature limits and for pressurized thermal shock evaluations are dependent on the reference temperature for nil-ductility transition (RTNDT). The original method to determine the RTNDT was formulated more than 20 yr ago when Section III of the ASME Code was adopted. At that time, there were insufficient data to judge whether some of the weld metals used in reactor vessel fabrication were unsuitable for this procedure. Presently, this causes a compliance problem for some weld metals used in nuclear reactor vessels, whereas there is no technical problem in meeting required safety margins. The RTNDT is a parameter to index degrees of irradiation embrittlement to adjust the Code reference fracture toughness curves to represent the actual degraded fracture toughness at a given fluence of a reactor vessel beltline region. When there is a problem determining RTNDT value for unirradiated material where Charpy transition temperature is the dominating criterion, an alternative RTNDT based solely on a drop-weight test was investigated for some of the weld metals. Using a new test method for fracture toughness in the transition range (ASTM, 1993), a fracture toughness curve was directly generated from a set of compact tension test data and used for validating the nil-ductility temperature (TNDT) from drop-weight test data as the sole mean for determining initial RTNDT value.

Role of irradiated materials data in performing reactor vessel fracture toughness analysis

The role of surveillance materials data in the various reactor vessel fracture toughness analyses required to insure that the vessels operate in accordance with defined regulations is often overlooked, because of the apparent prescriptive approaches that are used to perform the evaluations. These procedures are normally so conservative that, as the reactor vessel ages, the restrictions severely limit plant operation. A unique and often unrecognized feature of the regulatory guide provides for the use of surveillance data. This feature permits the plant to, obtain credit for having a viable surveillance program. The role of irradiated materials data is significant in the evaluation of three phases of reactor vessel operation or evaluation. These phases are as follows: normal operating pressure-temperature limits; evaluation of pressurized thermal shock; the case of a low Charpy upper shelf energy. The significance of having reliable irradiated fracture toughness data and their role in each of these evaluations are discussed. Data on irradiated material fracture toughness properties from reactor vessel surveillance programs improve the operating limitations of reactor vessels. These data permit the reduction of conservatisms applied when material-specific data are not available. The rewards of greater operating flexibility are sufficient that they encourage plant operators to maintain the optimum surveillance program for their reactor vessel.

Nuclear reactor pressure vessel-specific flaw distribution development

Vessel intergrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of U.S. vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired U.S. PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss also the development and application of the methodology and the impact to future vessel integrity analyses.

Reactor pressure vessel functionality issues and the development of a low temperature overpressurization transient limit for precluding crack initiation

Attainment of the design service life of the reactor pressure vessel (RPV) has become a difficult goal for some pressurized water reactors (PWRs) and consequently life extension may be precluded as an option. The difficulty has emerged due to the embrittlement of the pressure vessel steel from irradiation, and affects both reactor vessel integrity and functionality issues. Since the functional life of the reactor vessel is dependent upon the ability to adhere to the RPV pressure-temperature ( P-T) limits and the low temperature overpressure protection requirements, these normal operation constraints become quite important and control the reactor coolant system (RCS) operating window. This window is anticipated to close for many PWRs during their design life, preventing effective operation of the RCS and the RPV. To address this problem, the margin to incipient crack initiation associated with ASME Code Section III, Appendix G, fracture mechanics procedures has been evaluated and alternate brittle fracture limits for low temperature overpressure transients have been developed. These limits permit higher pressures for transient conditions than those associated with the normal operation P-T limits of Appendix G. The P-T limits for low temperature transient conditions have been developed, using the principles of linear elastic fracture mechanics. The stress intensity factors, K I have been developed through the use of a superposition technique with influence coefficients. The low temperature transient P-T limits preclude crack initiation for a range of postulated defect sizes and account for actual mechanical and thermal loadings, including those neglected in the simplified ASME Code procedures. This paper chronicles the RPV functionality issue along with the development of a low temperature overpressurization transient P-T limit for resolution of this operational problem.