ML19291C144
| ML19291C144 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1979 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19291C143 | List: |
| References | |
| REF-GTECI-A-11, REF-GTECI-RV, TASK-A-11, TASK-OR NUDOCS 8001230020 | |
| Download: ML19291C144 (12) | |
Text
,
Task A-ll Rev. 2 REACTOR VESSEL MATERIALS TOUGHNESS Lead NRR Organization:
Division of Operating Reactors (00R)
Lead Supervisor:
Darrell G. Eisenhut, Acting Director 2
Division of Operating Reactors Task Manager:
Richard Johnson, EB/ DOR Applicability:
All Reactors Projected Completion Date:
December 1979 l2 1791 103 8 0012 3 0 D2
Task A-11 Rev. No. 2 December 1979 1.
DESCRIPTION OF PROBLEM Because the possibility of failure of nuclear reactor pressure ves-sels designed to the ASME Boiler and Pressure Vessel Code is remote, the design of nuclear facilities does not provide protection against reactor vessel failure.
Prevention of reactor vessel failure depends primarily on maintaining the reactor vessel material fracture tough-ness at levels that will resist brittle fracture during plant opera-tion. At service times and operating conditions typical of current operating plants, reactor vessel fracture toughness properties pro-vide adequate margins of safety against vessel failure; however, as plants accumulate more and more service time, neutron irradiation reduces the material fracture toughness and initial safety margins.
Results from reactor vessel surveillance programs indicate that up to approximately 20 operating PWRs will have beltline materials with marginal toughness, relative to the requirements of Appendices G and H of 10 CFR Part 50, after comparatively short (approximately 10 EFPY) periods of operation. The specific requirement which may be violated is that of paragraph V.8, Appendix G, 10 CFR Part 50.
For vessels which fail to satisfy that requirement, paragraph V.C.3, Appendix G, 10 CFR Part 50, must be satisfied (along with the rest of V.C); that is, perform an analysis which demonstrates the existance of adequate operational safety margins against fracture.
For plants currently under licensing review, reactor vessels generally have acceptat;le fracture toughness. However, a few plants under licensing review have reactor vessels that have been identified as having the potential for marginal fracture toughness within their design life; 2
these vessels will have to be reevaluated in the light of the new criteria for long term acceptability.
The fundamental goal of Task A-11 is to provide an engineering method to assess the safety margin for failure prevention in nuclear reactor pressure vessels.
The method will employ the most advanced fracture mechanics concepts pressently available. Although linear elastic fracture mechanics analyses may be applicable at low temperatures, the amount of crack tip plastic deformation accompanying fracture at high temperature will be relatively large, even in pressure vessel steels of low toughness.
Therefore safety will be evaluated by comparing some measure of fracture resistance to a structural parameter, both being based on elastic plastic fracture mechanics concepts.
The concepts set forth in NUREG-0311, "A Treatment of the Subject of Tearing Instability," will be utilized to develop the required engineering method. Adequate margin will require that the structural parameter remain sufficiently below the measure of fracture resistance but the quantitative relationship may depend on the reactor plant conditions.
For example, a much larger margin would be required for normal / upset conditions than for low probability accident events.
1791 104 A-11/1
Task A-11 Rev. No. 2 December 1979 2.
PLAN FOR PROBLEM RESOLUTION The determination of appropriate licensing criteria for low toughness 2
reactor vessel materials and the evaluation of material degradation resulting from neutron irradiation demands an interdisciplinary effort encompassing several aspects of materials and fracture technology.
The plan for development of suitable licensing criteria for low toughness reactor vessel materials, including the effects of neutron irradiation damage, includes the following tasks.
A.
Identify ard measure the mechanical properties which control tearing instability types of fractures.
B.
Develop a method for analyzing structural members that incorpo-rates postulated flaws, under conditions which could lead to tearing instability fractures.
C.
Using the results from Sub-tasks A and B, define reactor vessel safety criteria to avoid failure by tearing instability fracture, to supplement existing criteria for other failure modes.
D.
Evaluate the feasibility of in place reactor vessel annealing to regain toughness.
E.
Evaluate actions which could lessen the severity of actual neutron radiation damage or improve the accuracy of calculations of such damage.
F.
Establish a computer information system for storage and retrieval of reactor pressure vessel materials data.
Each sub-task is discussed briefly in the rest of this section.
A.
Evaluate Material Fracture Resistance The measurement of fracture toughness for reactor vessel and other materials at temperatures corresponding to the upper shelf region is complicated by the presence of significant 2
pre-fracture plastic ficw.
Current toughness testing methods based on linear elastic fracture mechanics are not adequate to account for plastic flow.
New toughness testing techniques have been developed to allow evaluation of low toughness in 2
reactor vessel materials for normal, upset and accident conditions.
It is widely recognized that the J-integral provides a valid, general, solution to the problem of crack tip singularity fields under large-scale yielding, even up to fully plastic conditions for some geometries. Moreover, J has been shown 7
toprovideagoodindicationofsmall-scalecfackextension, 1/;l 103
,n.
A-11/2
Task A-ll Rev. No. 2 December 1979 although the ASTM has not yet established a standard test method for its measurement. More advanced work at Washington University, St. Louis, Missouri, under NRC funding, has resulted in the development of the tearing modulus, T, which is proportional to dJ/da. An experimental method, developed uader the Office of Nuclear Regulatory Research funding apparently can be used routinely to provide curves of J = f (Aa), the so-called J-R curves. From such data, both J and T TheformerhasprovenadequateIfagenN}canbedetermined.
fracture parameter; the latter provides a criterion fc" tearing instability where a large value of T indicates ductile tearing and a small valueindicatesMNfracture.
The goal of this sub-task is to provide the relevant materials mechanical property data for the evaluation of reactor vessel margin against fracture at temperatures above the ductile-brittle transition (beyond the range of linear elastic fracture mechanics applicability). Task A-ll will use data provided by the RES (NRC) - funded HSST Program which will include the effects of material condition, temperature and neutrun radiation.
2 8.
Develop Structural Analysis Methods Application of the tearing modulus concept to reactor vessel failure evaluation requires the development of a method for determining load carrying capacity.
Factors to be included in the analytical method must include the following.
The geometry of the component must be a basic consideration, including postulated flaw size, shape and orientation, in a parametric way.
Crack initiation and propagation will be characterized by parameters.
Loading I
J-integral and tearing modulus, T conditionswillincludetimedepe80Nc,eandtheroleofstructural compliance.
Temperature is a consideration to the extent that the instability analysis is applicable only above the ductile-brittle fracture mode transition. At relatively low temperatures the well-developed linear elastic fracture mechanics methods will be applicable.
The problem to be faced when considering reactor pressure vessel welds of marginal toughness is that neutron radiation can decrease the toughness, as represented by the Charpy upper shelf energy, below that required by current regulations.
Because of the dominant role of radiation-induced embrittlement, the elastic plastic response of the reactor pressure vessel beltline region will be controlling and will be the calculation used for the purpose of meeting the Task A-11 goal.
Completion of this sub-task will depend heavily on the results of the program being managed by Sandia, Albuquerque, with funds 1791 106 A-11/3
Task A-11 Rev. No. 2 December 1979 from DOE /NPD.
The principal contract in the program is with Washington University, St. Louis, Missouri, where the analytical method is being developed.
Additional help is available from PWR NSSS vendors
- through separate contracts placed by Sandia with each.
This sub-task will provide elastic plastic fracture mechanics formulations, applicable to reactor pressure vessel beltline regions, with which relevant structural parameters can be calculated for comparison to material properties (sub-task A) in order to evaluate failure margins.
C.
Define Safety Criteria To ensure adequate margins against failure for plants with marginal toughness materials in the reactor vessel beltline region, it will be necessary to establish suitable safety criteria for the vessels which fail to. satisfy the requirements of Section V.B Appendix G, 10 CFR Part 50.
Thr solution is to employ the elastic plastic fracture concepts set forth in NUREG-0311. The relevant materials mechanical properties will be those developed in sub-task A, above.
The reactor vessel beltline region will be analyzed with the methods developed in sub-task B, above.
The material parameters, such as J and 2
T can be compared to the structural parameters, su as J aMtt Fract698j.
Comparison, as was done in the report "A Preliminary Analysis on the Integrity of HSST Intermediate Test Vessels" by A. Zahoor, P. C. Paris and M. P. Gomez, is expected to show that crack extension occurs when J is the order of JIc and that the fracture mode depends on the relative values of T
and T T*atl to vaiO8d w(where fast fracture can be avoided by keeping ell below T This sub-t sk will provide m898I realisticcriteriaforENlu)a.
ting vessel fracture margins under normal, upset or faulted conditions at higher temperatures that the currently available linear elastic fracture mechanics.
The required margin of safety will depend on analyses of available fracture data (such as the HSST vessels) and on the severity of the given operating conditions.
D.
Evaluate Vessel Annealing Feasibility Thermal annealing to recover the toughness lost by neutron radiation was recognized as a theoretically possible method to regain toughness margins.
Studies are underway through contracts funded by the NRC and EPRI.
The feasibil hy studies will assess the practicality of reactor vessel recovery annealing.
Engineering guidance will be developed to help licensees determine the relative merits of vessel annealing to regain toughness.
" Namely: Babcock and Wilcox, Combustion Engineering and Westinghouse.
1791 107
- 3. u,,
Task A-11 Rev. No. 2 December 1979 E.
Radiation Damage Abatement The root cause of the reactor vessel toughness problem is neutron radiation. There are at least three aspects of neutron radiation which will be examined to determine their potential for reducing the severity of pressure vessel embrittlement or improving the accuracy of embrittlement calculation.
The thrust of this sub-task is to determine the amount of decrease in calculated mechanical property degradation which could be attained, while maintaining safety margins, by more exact neutron radiation calculations and to evaluate the potential for mitigating the problem through minor design changes.
(1) The neutron fluence through the vessel wall is calculated.
Some conservatism is purposely put into the calculations.
However, for marginal material, small decreases in calculted fluence could delay the point in time when the current code limits would be violated and, in some cases, could eliminate ^the problem altogether. The Office of Nuclear Regulatory Research, NRC, has an ongoing program which includes evaluation of neutron flux calculations and measurements. Although the program will not be completed within the term of Task A-11, early results may be used to assess the accuracy and margin of conservatism of vessel embrittlement calculations.
(2) Pre-service estimates of changes in reactor pressure vessel mechanical properties per unit fluence are based on relevant data, including those from test reactor experiments.
Vessel surveillance programs, required by 10 CFR Part 50, provide closer approximation from encapsulated specimens close to the vessel wall in the same reactor environment.
Surveillance data, as well as some long-term basic radiation experiments, can be used to modify relationships between fluence, inferred from calculation and measurement, and mechanical properties so that the predicted changes will be more realistic.
However, test reactor neutron radiation is significantly different from that through the vessel wall, particalarly with respect to dose rate and spectrum.
The extent to which such results are applicable to vessel steels with marginal toughness will be examined as part of this sub-task.
(3) To the extent that neutron fluence reductions can signifi-cantly reduce the rate of embrittlement, thereby delaying the advent of code violation, it is worthwhile to consider actions which would diminish the actual flux at the vessel.
Shielding, for example, might be inserted between the core and the vessel. Another possibility being considered by A-11/5
\\]0\\
Task A-ll Rev. No. 2 December 1979 some European operators is replacing corner fuel assemblies with dummies thus reducing the azimuthal neutron peaks.
2 F.
Establish a Vessel Data Information System Because of the large number of possible combinations of reactor vessel and surveillance materials and the large number of variables involved in evaluating these materials, it is neces-sary to develop an information system for the storage and retrieval of these data. This system will be utilized partic-ularly to maintain up-to-date, accurate data for the generic and plant specific evaluation of operating facilities.
This sub-task is part of a program funded by 00E/NPD, managed by 2
Sandia, Albuquerque, and is essentially complete.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION OF TASK As discussed in Section 1, the safety issue addressed by this task 2
is the reduction of reactor vessel material fracture toughness as a result of neutron irradiation. The operational temperature range includes the transition temperature region, where material toughness increases significantly with increasing temperature, and the upper shelf temperature region, where material toughness reaches a rela-tively constant maximum value. The task will develop licensing criteria to ensure that adequate margins of safety, relative to flaw-induced fracture, are maintained during normal operating and postulated accident conditions for reactor vessels containing belt-line (that part of the reactor vessel directly opposite the core) material with reduced toughness after prolonged irradiation.
For most plants now in the licensing process, current criteria, 2
together with the materials currently employed, are adequate to ensure suitable safety margins for the reactor vessels thoughout their design lives.
For currently operating plants, and for several plants in late stages of licensing that may have marginal toughness 2
materials, the safety margins required by Appendix G to 10 CFR Part 50 in the transition temperature region are, or will be, main-tained during normal operating conditions by appropriate shifts in the operating pressure-temperature limitation. Various analyses of accident conditions indicate that adequate material toughness in the transition temperature region will continue to be available to 2
ensure adequate safety margins for time periods significantly in excess of that required to complete this task.
A-11/6 1791 109
Task A-Il Rev. No. 2 December 1979 A few PWRs have reactor vessel beltline materials whose upper shelf l 2 energies may fall below levels required by Appendices G and H to 10 CFR Part 50 within the next few years. An interim assessment
- was made of the safety margins with respect to flaw-induced fracture 2
for operating vessels with low upper shelf beltline materials. The evaluation indicated that adequate margins of safety can be maintained in the interim period prior to completing this task for the postulated stress and flaw conditions spe'cified in Appendix G to Section III of the ASME Code and required by Appendix G to 10 CFR Part 50.
I 2 Pending completion of this task, the safety margins required by Appendix G to 10 CFR Part 50 for operation in the transition temper-ature region can be maintained during normal operation by appropriate shifts of the operating pressure-temperature limits as dictated by the material surveillance program results and Regulatory Guide 1.99.
2 Initial analyses submitted by some NSSS vendors and our preliminary review indicate that adequate toughness margins can also be maintained in the transition region for postulated accident conditions for up to approximately 20 years of neutron irradiation, or significantly beyond completion of this task.
Appendix G to 10 CFR Part 50 requires licensees of those plants where the beltline material upper shelf energy is predicted to fall below 50 ft-lb to conduct a 100% volumetric examination of the low toughness beltline material. This examination provides added assur-ance that very large flaws are not present in the reactor vessel beltline region.
Should the results of this task indicate that in the future adequate l 2 margins of safety for the reactor vessels of operating plants cannot be demonstrated for both normal operation and postulated accident condi-2 tions, one or more of the following alternative measures can be taken.
1.
Reactor vessel annealing to regain material toughness in the beltline region.
l 2 2.
Increased beltline inspections using improved in-service inspection (ISI) techniques, as they become available with demonstrated required reliability, leading to a justified decrease in postulated flaw size.
3.
Modifications to the vessel internals or core design to modify 2
the neutron flux and reduce subsequent material degradation.
4.
System modifications io limit the severity of loading (stress levels) of the reactor vessel during postulated emergency or accident conditions.
- Memoranoum, V. S. Noonan to D. G. Eisenhut, " Reactor Vessels with Marginal Toughness Properties", July 19, 1979.
A-11/7 179 10
Task A-11 Rev. No. 2 December 1979 In summary, the staff considers that in the interim period the safety margins are adequate to ensure the safety of reactor vessels in currently operatir.g plants.
The current licensing criteria and the materials used for reactor vessel fabrication provide assurance that reactor vessels for those plants now in the licensing process also will have adequate margins of safety relative to flaw-induced failure.
Accordingly, we conclude that while the task is being performed, continued operation and plant licens'ing can proceed with reasonable assurance of protection to the health and safety of the public.
4.
NRR TECHNICAL ORGANIZATIONS INVOLVED A.
Engineering Branch, Division of Operating Reactors.
Has overall lead responsibility in the identification of relevant reactor vessel material in licensed plants, evaluation of operating experience with neutron irradiation damage, determination of tne associated degradation in reactor vessel material toughness and the evaluation and determination of an appropriate safety criterion for low toughness reactor vessel materials.
Manpower Estimates:
1.5 man years FY 1979; 0.5 man year FY 1980. l 2 B.
Materials Engineering Branch, Division of Systems Safety. Has lead respnnsibility for review of experimentally determined materials fracture resistance as a function of neutron radiation, 2
for developing the NRC position on in place reactor vessel annealing and for evaluation of information developed during the evaluation of material toughness in licensed facilities for possible inclusion into material toughness criteria currently used for facilities not yet licensed for operation, where appropriate.
Manpower Estimates: 0.2 man year FY 1979; 0.1 man year FY 1980.
2 C.
Reactor Safety Branch, Division of Operating Reactors. Has lead responsibility for review of neutron fluence calculation methods. Will advise EB/00R with respect to the application of the results from the RES radiation damage program to the problem of predicting reactor vessel damage and the advisability of recommending shielding or core modifications to mitigate neutron damage.
Manpower Estimates: 0.2 manyear FY 1979; 0.1 man year FY 1980.
D.
Environmental Projects Branch 2, Division of Site Safety and Environmental Analysis. Has lead responsibility for defining A-11/8 l79l
}$
Task A-11 Rev. No. 2 December 1979 licensing criteria related to effluent and personnel exposure control during reactor vessel annealing operations.
Manpower Estimates: 0.04 manyear FY 1979; 0.02 man year FY 1980.
l2 5.
TECHNICAL ASSISTANCE Technica' assistance from organizations catside the NRC will be required to complete Tasks A through F in Section 2, Plan for Problem 2
Resolution, i.e., all aspects of the Task Action Plan. The contractors assisting in these tasks are as follows:
A.
Contractor: Washington University (EB/ DOR)
Funds Required: $103.6K FY 1979; $7K FY 1980.
This program is directed specifically at Tasks 2-A, -B, and -C.
The results of the program will allow advanced fracture mechanics techniques to be used to establish a technical basis for NRC's dcvelopment of a suitable licensing criterion for low toughness materials. Associated with this is the determination of simplified analytical techniques to evaluate normal operating conditions, postulated accident conaitions and assistance in plant specific analyses.
B.
Contractor: Naval Research Laboratory (EB/00R, MTEB/ DSS)
Funds Required:
$30K FY 1979; $15K FY 1980.
2 This program will investigate neutron irradiation of reactor vessel steels and is directed specifically at Task 20, Evaluate Vessel Annealing Feasibility.
The results should provide improved means to quantitatively describe the effects of material microstructure, chemical composition, neutron spectra and dose rate and allow suitable evaluation, prediction and monitoring of irradiation damage to reactor vessel steels.
Included in this program is a study of the feasibility of in place annealing of reactor vessels to restore fracture toughness to levels that will provide adequate safety margins.
C.
Contractor: Brookhave'n National Laboratory (RS/00R)
Funds Required: $15K FY 1979; $5K FY 1980.
This program will provide independent neutron flux (fluence) calculations including the effects of core and structural configurations and energy spectra.
A-11/9
}79
Task A-11 Rev. No. 2 December 1979 6.
INTERACTION WITH OTHER OUTSIDE ORGANIZATIONS A.
Sandia (Albuquerque)/ DOE A program will be managed by the Light Water Reactor R&D Section, 2
Sandia, with funds provided by 00E/NPD. The program will have two goals:
(1) develop an information system to complete sub-task F; (2) develop.the analytical basis for evaluating the margin against pressure vessel fracture will be determined using elastic plastic analyses as described in sub-task 8.
B.
Licensees Intermittent interaction with licensees is expected for the purpose of obtaining required materials data.
C.
NSSS Vendors Some plant specific analyses have been conducted by the NSSS vendors.
Review of the portions of these analyses relevant to completion of the generic task will be required.
Some NSSS vendors have first-hand knowledge of fabrication and materials data relevant to low material toughness; review of these data will be required.
D.
EPRI EPRI is currently funding a number of programs related to reactor vessel materials toughness. These programs include studies for neutron irradiation damage of pressure vessel steels and the development of fundamental failure criteria based on elastic plastic fracture mechanics.
Interaction with EPRI to remain informed on the direction and results of these programs and to ensure that appropriate NRC licensing concerns are addressed will be required.
E.
ACRS This task is closely related to one of the generic items identified by the ACRS and, accordingly, will be coordinated with the Committee as the task progresses.
7.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES A.
Office of Nuclear Regulatory Research, Division of Reactor Safety Research, Metallurgy and Materials Branch.
A-11/10
) ?\\
Task A-11 Rev. No. 2 December 1979 RES is funding a major experimental research program (Heavy Section Steel Technology, HSST) through Oak Ridge National Laboratory to determine the fracture toughness of reactor vessel steels and the safety margins for reactor vessels. At the request of NRR, RES modified this progran to include materials with low toughness, representative of those at operating 2
facilities.
(Sub-task A)
At the request of NRR, RES is supporting a program to verify experimentally the application of the tearing stability concept as a failure criterion for beltline materials with marginal fracture toughness.
(Sub-task A) 2 RES initiated a comprehensive research program to experimentally validate neutron irradiation damage in pressure vessel steels and the associated calculational schemes used to predict radiation damage. This effort is to be part of an overall program being conducted in cooperation with research groups in the US and Europe.
(Sub-task E) l2 8.
Office of Standards Development, Division of Engineering Standards, Structures and Components Standards Branch.
SD is assisting NRR in the study of the effects of neutron irradiation and the evaluation of low toughness reactor vessel 2
steels.
(Sub-task C)
C.
Office of Management Information and Program Control, Division of Regulatory Information Systems, Proces3ing and Programming Branch.
MIPC provides assistance to NRR toward the goal of establishing a computer-based information system for the storage and retrieval 2
of materials surveillance data.
(Sub-task F) 8.
F0TENTIAL PROBLEMS All sub-tasks are well on the way toward completion, however, it is not clear at this time that the state of tne art is far enough advanced to allow completion of Sub-task 8 with the present allotment of resources.
Also, tfie amount of data which can be provided by sub-task A may be insufficient to provide an accurate evaluation of vessel material behavior.
A-ll/ll
)50
E. G. Case cc:
D. Eisenhut V. Noonan L. Shao J. Strosnider R. E. Johnson M. Aycock P. Kapo R. Klecker W. Hazelton J. Knight P. Check S. Pawlicki W. Regan J. Watt S. Varga P. Randall, OSD C. Serpan, RES R. Gamble P. Leech D. Sells D. Muller W. Anderson R. Vollmer B. Grimes I. Kirk D. Davis B. Morris EB Members NRC PDR Accession Unit ACRS (10 Copies)
)
0