1、Designation: E2956 14Standard Guide forMonitoring the Neutron Exposure of LWR Reactor PressureVessels1This standard is issued under the fixed designation E2956; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of last revis
2、ion. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.INTRODUCTIONLight Water Reactor (LWR) power plant safety analysis reports and subsequent neutron exposureparameter calculations for the reac
3、tor pressure vessel (RPV) wall and critical welds need to be verifiedusing modern codes and information from surveillance dosimetry. The location of critical weldsrelative to the axial and azimuthal fluence rate map should be taken into account, as well as changesin fuel loading during periods when
4、surveillance capsules are exposed and beyond to the end of thereactors operating license. For many reactors today this is a 60-year-long interval. In the nuclearindustry, there is active consideration and evaluation of an 80-year-long operating interval. Mostreactor surveillance programs were design
5、ed based on the guidance of Practice E185 with a 40-yearoperating life in mind. The Practice E185 surveillance programs are designed to select and irradiatethe RPV material test specimens. The dosimetry in the surveillance capsule is there primarily tomeasure the neutron fluence to which the capsule
6、s material specimens have been exposed.In addition, those programs were based on the operating assumptions in place at the time; typicallyannual out-in core loading patterns and base load operation at a fixed reactor power level. Reactoroperations have evolved so that low-leakage core loading patter
7、ns (L3P) are the norm as are 18- and24-month-long fuel cycles and reactor power up-ratings of up to 20 %. Many reactors have nowinstalled flux suppression features such as natural uranium fuel rods, full or part-length hafnium orB4C rods, or stainless steel rods to minimize the neutron exposure of c
8、ritical areas of the RPV. Suchdevelopments increase the need to comprehensively monitor the RPV accrued fluence through theextended operation period.This guide is intended to be used together with other Standards to provide best estimates of theneutron exposure and exposure rate (together with uncer
9、tainties) at positions at the inner diameter andwithin the pressure vessel wall of a light water reactor. Also provided will be estimates of gamma-rayexposure and exposure rates to interpret dosimetry sensor photo-reaction and other gamma-rayinduced effects. Information used to make these estimates
10、is obtained from coupled neutron-gammaray transport calculations and from neutron and gamma-ray sensors located in surveillance positionson the core side of the vessel and in the reactor cavity outside the vessel wall (1).2Benchmark fieldirradiations of similar monitors also provide valuable informa
11、tion used in the verification of theaccuracy of the calculations (1).Knowledge of the time-dependent relationship between exposure parameters at surveillancelocations and selected (r, , z) locations within the pressure vessel wall is required to allowdetermination of the time-dependent radiation dam
12、age to the RPV. The time dependency must beknown to allow proper accounting for complications due to burn-up, as well as changes in core loadingconfigurations (2-5). An estimate of the uncertainty in the neutron exposure parameter values atselected (r, , z) points in the vessel wall (1) is also need
13、ed to place an upper bound on the allowableoperating lifetime of the reactor vessel without remedial action (6-9). (See Guide E509.)Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States11. Scope1.1 This guide establishes the means and frequ
14、ency ofmonitoring the neutron exposure of the LWR reactor pressurevessel (including the extended beltline) throughout its operat-ing life.1.2 The physics-dosimetry relationships determined fromthis guide may be used to estimate reactor pressure vesseldamage through the application of Practice E693 a
15、nd GuideE900, using fast neutron fluence (E 1.0 MeV andE0.1MeV), displacements per atom dpa, or damage-function-correlated exposure parameters as independent exposure vari-ables. Supporting the application of these standards are theE853, E944, E1018, and E1005 standards, identified in 2.1.1.3 This s
16、tandard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.
17、1 ASTM Standards:3E170 Terminology Relating to Radiation Measurements andDosimetryE185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E509 Guide for I
18、n-Service Annealing of Light-Water Mod-erated Nuclear Reactor VesselsE693 Practice for Characterizing Neutron Exposures in Ironand Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706(ID)E844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)E853 Practice for
19、Analysis and Interpretation of Light-WaterReactor Surveillance ResultsE900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel Materials, E706 (IIF)E944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E1005 Test Metho
20、d for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E 706(IIIA)E1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E706 (IIB)E2005 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Neutron FieldsE2006 Guide for Ben
21、chmark Testing of Light Water ReactorCalculationsE2215 Practice for Evaluation of Surveillance Capsulesfrom Light-Water Moderated Nuclear Power Reactor Ves-sels2.2 American Society of Mechanical Engineers Standard:Boiler and Pressure Vessel Code, Sections III and XI42.3 Nuclear Regulatory Document:C
22、ode of Federal Regulations, Chapter 10, Part 50, AppendixA “General Design Criteria for Nuclear Power Plants,”Appendix G “Fracture Toughness Requirements,” andAppendix H Reactor Vessel Material Surveillance Pro-gram Requirements”53. Terminology3.1 Definitions for terms used in this guide are found i
23、nTerminology E170.4. Significance and Use4.1 Regulatory RequirementsThe USA Code of FederalRegulations (10CFR Part 50, Appendix H) requires the imple-mentation of a reactor vessel materials surveillance programfor all operating LWRs. Other countries have similar regula-tions. The purpose of the prog
24、ram is to (1) monitor changes inthe fracture toughness properties of ferritic materials in thereactor vessel beltline region resulting from exposure toneutron irradiation and the thermal environment, and (2) makeuse of the data obtained from surveillance programs to deter-mine the conditions under w
25、hich the vessel can be operatedwith adequate margins of safety throughout its service life.Practice E185, derived mechanical property data, and (r, , z)physics-dosimetry data (derived from the calculations andreactor cavity and surveillance capsule measurements (1) usingphysics-dosimetry standards)
26、can be used together with infor-mation in Guide E900 and Refs. 4, 10-17 to provide a relationbetween property degradation and neutron exposure, com-monly called a “trend curve.” To obtain this trend curve at allpoints in the pressure vessel wall requires that the selectedtrend curve be used together
27、 with the appropriate (r, , z)neutron field information derived by use of this guide toaccomplish the necessary interpolations and extrapolations inspace and time.4.2 Neutron Field CharacterizationThe tasks required tosatisfy the second part of the objective of 4.1 are complex andare summarized in P
28、ractice E853. In doing this, it is necessaryto describe the neutron field at selected (r, , z) points withinthe pressure vessel wall. The description can be either timedependent or time averaged over the reactor service period ofinterest. This description can best be obtained by combiningneutron tra
29、nsport calculations with plant measurements such asreactor cavity (ex-vessel) and surveillance capsule or RPVcladding (in-vessel) measurements, benchmark irradiations of1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility
30、of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Feb. 1, 2014. Published March 2014. DOI: 10.1520/E2956-14.2The boldface numbers in parentheses refer to the list of references appended tothis guide.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orc
31、ontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available from American Society of Mechanical Engineers (ASME), ASMEInternational Headquarters, Two Park Ave., New York, NY 10016-599
32、0, http:/www.asme.org.5Available from U.S. Government Printing Office Superintendent of Documents,732 N. Capitol St., NW, Mail Stop: SDE, Washington, DC 20401, http:/www.access.gpo.gov.E2956 142dosimeter sensor materials, and knowledge of the spatial corepower distribution, including the time depend
33、ence. Becausecore power distributions change with time, reactor cavity orsurveillance capsule measurements obtained early in plant lifemay not be representative of long-term reactor operation.Therefore, a simple normalization of neutron transport calcu-lations to dosimetry data from a given capsule
34、is unlikely togive a satisfactory solution to the problem over the full reactorlifetime. Guide E482 and Guide E944 provide detailed infor-mation related to the characterization of the neutron field forBWR and PWR power plants.4.3 Fracture Mechanics AnalysisCurrently, operatinglimitations for normal
35、heat up and cool down transientsimposed on the reactor pressure vessel are based on the fracturemechanics techniques outlined in the ASME Boiler and Pres-sure Vessel Code. This code requires the assumption of thepresence of a surface flaw of depth equal to one fourth of thepressure vessel thickness.
36、 In addition, the fracture mechanicsanalysis of accident-induced transients (Pressurized ThermalShock, (PTS) may involve evaluating the effect of flaws ofvarying depth within the vessel wall (4). Thus, information isrequired regarding the distribution of neutron exposure and thecorresponding radiati
37、on damage within the pressure vessel,both in space and time (4). In this regard, Practice E185provides guidelines for designing a minimum surveillanceprogram, selecting materials, and evaluating metallurgicalspecimen test results for BWR and PWR power plants. PracticeE2215 covers the evaluation of t
38、est specimens and dosimetryfrom LWR surveillance capsules.4.4 Neutron Spectral Effects and DPAAnalysis of theneutron fields of operating power reactors has shown that theneutron spectral shape changes with radial depth into thepressure vessel wall (2, 3). The ratio of dpa/t (where is thefast (E 1.0
39、MeV) neutron fluence rate and t is the time thatthe material was exposed to an average fluence rate) changesby factors of the order of 2.0/1.0 in traversing from the inner tothe outer radius.Although dpa, since it includes a more detailedmodeling of the displacement phenomenon, should theoreti-cally
40、 provide a better correlation with property degradationthan fluence (E 1.0 MeV) (1, 18), this topic is stillcontroversial and the available experimental data does notprovide clear guidance (18, 19). Thus it is recommended tocalculate and report both quantities; see Practice E853 andPractice E693.4.5
41、 In-Vessel Surveillance Programs:4.5.1 The neutron dosimetry monitors used in reactor vesselsurveillance capsules provide measurements of the neutronfluence and fluence rate at single points on the core midplanewithin the reactor, and near the vessel wall; that is, at thesurveillance capsule locatio
42、ns (1). In actual practice, thesurveillance capsules may be located within the reactor at anazimuthal position that differs from that associated with themaximum neutron exposure (or that differs from the azimuthaland axial location of the assumed flaw); and at a radial positiona few centimeters or m
43、ore from the flaw and the pressure vesselwall (4, 5). Although the surveillance capsule dosimetry doesprovide points for normalization of the neutron physics trans-port calculations, it is still necessary to use analytical methodsthat provide an accurate representation of the spatial variation(axial
44、, radial and azimuthal) of the neutron fluence (refer toGuide E482). It is also necessary to use other measurements toconfirm the spatial distribution of RPV neutron exposure.4.5.2 Given that surveillance capsules are located radiallycloser to the core than the surface of the RPV, they may beshifted
45、 azimuthally away from the peak exposure location inorder to limit the magnitude of the surveillance capsule leadfactor. The lead factor is defined as the ratio of the fast neutronfluence at the center of the surveillance capsule to the peak fastneutron fluence at the clad base metal interface of th
46、e RPV.One adverse effect of this azimuthal shift away from the peakis that the surveillance capsule dosimetry does not “see” thepart of the core that produces the peak exposure of the reactorvessel.As a result, the surveillance capsule is unable to monitorthe effect of changes in the core power dist
47、ribution that aremade to reduce the peak RPV neutron exposure. Anotheradverse effect is that with larger lead factors, the capsules arerapidly exposed to a high neutron fluence. For example, with alead factor of five, a surveillance capsule will receive anexposure in as little as 12 years that is eq
48、uivalent to what thereactor pressure vessel peak may see in 60 years of operation.Practices E185 and E2215 suggest not exceeding twice themaximum design fluence (MDF) or twice the end-of-licensefluence (EOLF). In this example, this would require withdraw-ing any remaining surveillance capsules after
49、 24 years ofoperation. Thus, without taking other steps, the reactor wouldbe operated for the remaining 36 years (of a 60-year life) withno dosimetry present.4.5.3 New or replacement surveillance capsules shouldrecognize and correct operating deficiencies by using improvedcapsule dosimetry. For example, for one class of PWR, thecopper wire is cadmium shielded to minimize interferencefrom trace amounts of cobalt. In about one third of themeasurements the copper has become incorporated into thecadmium preventing separation and further processing. Asimple solut
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