ASME STP-NU-010-2007 REGULATORY SAFETY ISSUES IN THE STRUCTURAL DESIGN CRITERIA OF ASME SECTION III SUBSECTION NH For Very High Temperatures for VHTR & Gen IV《VHTR & GEN IV反应堆用Sec .pdf

1、STP-NU-010REGULATORY SAFETY ISSUESIN THE STRUCTURAL DESIGNCRITERIA OF ASME SECTION IIISUBSECTION NHFor Very High Temperatures for VHTR 2) describes how Subsection NH addresses these issues; and 3) identifies the needs for additional criteria to cover unresolved safety concerns for very-high-temperat

2、ure service. The report also contains a description of the high-temperature structural integrity safety concerns raised by the U.S. Nuclear Regulatory Commission (NRC) and the Advisory Committee on Reactor Safeguards (ACRS) and how these issues are addressed in Subsection NH of the ASME Code. Regula

3、tory Issues in NH Structural Design Criteria STP-NU-010 1 1 SUMMARY The U.S. Nuclear Regulatory Commission (NRC) and Advisory Committee on Reactor Safeguards (ACRS) issues which were raised in conjunction with the licensing of the Clinch River Breeder Reactor (CRBR) provide the best early indication

4、 of regulatory licensing issues for high-temperature reactors. A construction permit for CRBR was supported by the ACRS with the stipulation that numerous ACRS/NRC technical issues be resolved prior to requesting an operating license. The research and development (R Coolant impurities and crevice co

5、ncentration impacts; Metal carburization, decarburization and oxidation; Sensitization of austenitic steels; Alloy aging behavior at elevated temperatures; The adequacy of in-service inspection; and The applicability/adequacy of the ASME Code database. Appendix B summarizes the current NRC licensing

6、 issues for the structural design of VHTR and Gen IV systems. In order to resolve these issues, Subsection NH of Section III of the Code and the Code Cases for elevated- temperature design require further development. The metal temperature limits of the Code need to be extended from 760C (1400F) to

7、at least 900C (1652F). The design lifetime limit of 34 years needs to be extended to 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered

8、. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction

9、 between the time-independent and time-dependent material response. The manner in which NRC licensing issues for the structural design of VHTR and Gen IV systems are addressed in the current ASME Subsection NH and Code Cases is described in section 5 of this report. The materials creep behavior, cre

10、ep fatigue and environmental effects are addressed in Subsection NH and Code Cases largely in terms of design criteria and allowable stress and strain values. The detailed material properties needed for cyclic finite element creep design analyses are generally not provided in the Code. The minimum s

11、trength properties given in the Code are used as anchor values for the more comprehensive material suppliers average properties. The NRC perspective is that the Code and/or Code Cases currently do not adequately cover the material behavior under cyclic loads in the creep regime, and creep fatiguecre

12、ep rupture damage interaction effects. Regulatory Issues in NH Structural Design Criteria STP-NU-010 3 Subsection NH has rules for the design of welded joints separated into categories A through D. The permissible types of welded joints and their dimensional requirements are specified. Paragraph NH-

13、3353 provides analysis requirements for the design and location of all pressure retaining welds operating at temperatures where creep effects are significant. Reduction factors for creep stress rupture are given as a function of time and temperature. Permissible weld metals are limited and special e

14、xamination requirements are imposed. Probably the most restrictive Subsection NH requirements for welds are that the inelastic accumulated strains are limited to one-half the allowable strain limits for the base metal. This has forced designers to keep welds out of high-stress areas. The allowable f

15、atigue at weldments is limited to one-half the design cycles allowed for the base metal. The allowable creep rupture damage at weldments is limited in NH by requiring that the rupture strength be reduced by the weld strength reduction factor when determining the time-to-rupture. The Code also impose

16、s additional examination requirements on Category A thru D welded joints. The adequacy of these and other Code weldment structural design requirements has been questioned by the NRC, even for the temperatures currently covered, which are lower than the VHTR and Gen IV High-Temperature Systems. Secti

17、on 6 of this report describes the material models, design criteria and analyses methods which NRC has indicated are remaining needs in the ASME Code to cover Regulatory Issues for Very High Temperature Service. The Code technical committees involved are listed for each of these needs: 1. Material cy

18、clic creep behavior, creep rupturecreep fatigue interaction and environmental effects. 2. The structural integrity of welds 3. The development of extended simplified design analysis methods (to avoid dependence on “black box” finite element analysis (FEA) for cyclic creep) 4. Test verification of 1,

19、 2 and 3 The NRC is currently expanding its staff to deal with their increased licensing workload for Gen III reactors as well as to address Gen IV technical licensing issues. They have expressed concerns about the validity of extending the current technology of Subsection NH to much higher temperat

20、ures, and see the need to resolve new corrosion and structural integrity issues for the materials to be used for very-high-temperature applications. Appendix B gives the current (May 2, 2007) NRC Draft for Review, Further Analysis of Elevated Temperature Structural Integrity (Licensing) Issues. STP-

21、NU-010 Regulatory Issues in NH Structural Design Criteria 4 2 INTRODUCTION The objective of this report is to identify issues relevant to ASME Section III, Subsection NH 1, and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very-High-Temperature Gas Reactor concepts such

22、 as those of PBMR Ltd., Areva and General Atomics); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for ele

23、vated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Bree

24、der Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTG

25、Rs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This report describes: 1. NRC and ACRS safety concerns raised during the licensing process of CRBR 2. How some of these issues are addressed by the current Subsection NH of the ASME Code 3. The material m

26、odels, design criteria and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very-high-temperature service Regulatory Issues in NH Structural Design Criteria STP-NU-010 5 3 NRC AND ACRS SAFETY ISSUES IN LICENSING REVIEW OF CRBR This sect

27、ion describes NRC staff and ACRS safety concerns with regard to the elevated-temperature structural design of LMFBR systems, related to licensing of the CRBR that took place during the late 1970s and early 1980s. The ACRS has statutory responsibilities as described in the Atomic Energy Act of 1954,

28、as amended. The ACRS reviews and advises the Commission with regard to the licensing and operation of production and utilization facilities and related safety issues, the adequacy of proposed reactor safety standards, technical and policy issues related to the licensing of evolutionary and passive p

29、lant designs, and other matters referred to it by the Commission. 3.1 Elevated-Temperature Design and Operating Licensing Conditions In order to assess the relevance of issues identified by the NRC and ACRS licensing reviews of CRBR to the structural design of VHTR and GEN IV systems, it is necessar

30、y to consider the specific design and operating conditions of the CRBR. The Clinch River Breeder Reactor Plant was designed to demonstrate that a liquid-metal fast breeder reactor can operate safely and reliably in an electric utility system. The plant was designed as a 350 MWe, three-loop system to

31、 be located in the Tennessee Valley Authority system at a site on the Clinch River near Oak Ridge, Tennessee. With a reactor vessel outlet temperature of 995F (535C) it was necessary in the structural design of the plant to take account of loading conditions and component response unique to elevated

32、-temperature service-enhanced thermal transients and gradients, nonlinear deformation and creep of materials, and time-dependent failure modes. With a design life of 30 years it was necessary to take account of material degradation effects due to sustained load and environment, geometry change due t

33、o creep, and a potential for loss of function. Since LMFBR systems operate at low pressure, the sodium containing components (reactor vessel, tanks, piping, heat exchangers, steam generators, pumps, and valves) are relatively thin-walled. Besides steady loads due to pressure, thermal expansion, and

34、dead weight, there are cyclic loads due to thermal transients, pressure changes, and seismic events. Thermally induced stresses become more significant in elevated-temperature systems, so additional attention must be paid to elastic follow-up, strain concentration, and geometrical instability. In co

35、ntrast with low-temperature design where the response is time-independent, cyclic loads combine with elevated-temperature, time-dependent material behavior making it necessary to follow the actual load history through time and to predict response as a function of time. The ordering of events, as wel

36、l as the time between events, may have a significant effect on response. 3.2 Structural Integrity Evaluation Approach for Licensing 3.2.1 Modes of Failure to Consider Elevated-temperature CRBR systems and components were designed to meet the limits of the ASME Boiler and Pressure Vessel Code, Sectio

37、n III, Case N-47 (1981) 2, the forerunner of Subsection NH, which applies for ferritic steels at temperatures above 700F (371C) and for austenitic stainless steels above 800F (427C). Failure is prevented by: (1) identifying each possible failure mode, (2) determining the damage criterion for each fa

38、ilure mode, and (3) establishing design rules that appropriately separate design limits from initiation of failure. Other rules rely on control of geometry, design rules to specify details, and design factors based on experience to avoid failure, but do not treat each failure mode explicitly. Case N

39、-47 is based primarily on design by analysis since it is not possible to develop simple, generally applicable formulas to represent the time-dependent response of complex structures for a 30-year life. However, it did include a number of simplified limits and bounding methods. The latter STP-NU-010

40、Regulatory Issues in NH Structural Design Criteria 6 were based on elastic and short-time plastic analyses which, although generally conservative, if satisfied, could avoid more detailed time-dependent, inelastic and creep analyses. The cost of analysis was a consideration. The modes of structural f

41、ailure considered in CRBR design include: Ductile rupture from short-term loads Creep rupture from long-term loads Creep-fatigue failure Gross distortion due to incremental collapse and ratcheting Loss of function due to excessive deformation Buckling due to short-term loads Creep buckling due to lo

42、ng-term loads 3.2.2 Stress Categories In Code Case N-47, stresses and strains are categorized as primary (P), secondary (Q) or peak (F), and in applying the limits, distinction is made between two types of quantities: (1) load-controlled and (2) deformation-controlled. The load-controlled quantities

43、 result from equilibrium with applied loads during plant operation. Primary stress intensities are load-controlled quantities. Deformation-controlled quantities are stresses, strains and deformations that result from deflection and/or strain compatibility. These quantities generally vary both with t

44、ime and applied loads, and creep effects may be a major influence. Thus, accurate analytical evaluation of deformation-controlled quantities generally requires inelastic stress analysis when creep effects are significant. 3.2.3 Material Representation Modeling of time-dependent material behavior in

45、multidimensional states of stress is fundamental to the accurate prediction of component response to service loads and to comparison with design limits. For CRBR the material models (constitutive equations) were developed by the Oak Ridge National Laboratory and are described in Reactor Development

46、(2) the effects of large variations of material properties within the weldment on creep-fatigue and creep-rupture damage; and (3) the effects of time rate, cycle rate and hold time on the propagation of long shallow cracks in the HAZ of a weldment. The NRC was also concerned about creep enhancement

47、of crack growth in a cracked weldment, specifically, enhanced creep in the remaining uncracked wall caused by residual stress and thermal cycling, and effects of creep on stability of the remaining uncracked wall ligament. The NRC felt that as a minimum these effects must be considered and quantitat

48、ively evaluated in order to determine the safety margins of weldments in elevated-temperature components. The basic concerns identified by NRC were: Evaluate potential for premature crack initiation at weldments due to thermal fatigue, residual stresses and damage caused by the welding process. Conf

49、irm adequacy of creep-rupture and creep-fatigue damage evaluation procedures at weldments. Assess growth behavior of cracks in the heat affected zone of weldments. Evaluate consequences of enhanced creep in uncracked ligaments. Assess stability of uncracked ligaments for creep conditions. Define effects of long-term elevated-temperature service on crack initiation. Evaluate effects of loading sequence on creep-fatigue behavior. It was required that these investigations be completed prior to issuance of a plant operating license.