ML19331C026
| ML19331C026 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1980 |
| From: | Cheng C, Gamble R, Taboada A Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0313, NUREG-0313-R01, NUREG-313, NUREG-313-R1, NUDOCS 8008140059 | |
| Download: ML19331C026 (38) | |
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ERRATA SHEET FOR Technical Report on Material Selection and Processing Guidelines f~
BWR Coolant Pressure Boundary Piping NUREG-0313, Rev. 1 The first single asterisk (*) appearing on page 5 refers to the footnote that was inadvertently placed I
on page 4 The two footnotes (*, **) appearing on page 5 are referenced in the last line of text on page 5.
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1 NUREG-0313 Rev.1 Technical Report on Material i Selection and Processing Guidelines i for BWR Coolant Pressure l Boundary Piping Resolution of Generic Technical Activity A-42
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Manuscript Completed: July 1980 Date Published: July 1980 l
C. Y. Cheng, R. M. Gamble, A. Taboada, B. Turovlin Divi:lon of Licensing Division of Engineering Offica of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commissinn Washington, D. C. 20555
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ABSTRACT This report updates and supercedes the NRC technical positions established in NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," published in July 1977.
This report sets forth the NRC staff's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1, 2, and 3 pressure boundary piping and safe ends.
For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented.
iii
CONTENTS P, age ABSTRACT.................................................................
iii I.
INTRODUCTION........................................................
1 II.
IMPLEMENTATION OF MATERIAL SELECTION, TESTING AND PROCESSING GUIDELINES..........................................................
4 III.
SUMMARY
OF ACCEPTABLE METHODS TO MINIMIZE CRACK SUSCEPTIBILITY -
MATERIAL SELECTION, TESTING, AND PROCESSING GUIDELINES..............
4 A.
Selection of Materials.........................................
4 1.
Corrosion-Resistant Materials.............................
4 2.
Corrosion-Resistant Safe Ends and Thermal Sleeves.........
5 B.
Testing of Materials...........................................
5 C.
Processing of Materials........................................
6 IV.
INSERVICE INSPECTION AND LEAK DETECTION REQUIREMENTS FOR BWRs WITH VARYING DEGREES OF CONFORMANCE TO MATERIAL SELECTION, TESTING, AND PROCESSING GUIDELINES...............................................
6 V.
GENERAL RECOMMENDATIONS.............................................
12 VI.
REFERENCES..........................................................
14 APPENDICES APPENDIX A - Task Action Plan A Pipe Cracks in Boiling Water Reactors....................................................
A-1 APPENDIX B - NRC Notice in the Federal Register Requesting Public Comment on NUREG-0531..................
B-1 APPENDIX C - Summary of Public Comments on NUREG-0531..................
C-1 v
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TECHNICAL REPORT ON MATERIAL SELECTION AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE BOUNDARY PIPING I.
INTRODUCTION This report is an update of the NRC technical position defined in NUREG-0313,
" Technical Report on Material Se16ction and Processing Guidelines for BWR Coolant Pressure Boundary Piping," July 1977 (Ref.1).
It constitutes the resolution of subtask C-1 and partial resolution of subtask C-2 of Generic Task No. A-42, " Pipe Cracks in Boiling Water Reactors." (Task Action Plan A-42 is attached as Appendix A.) This report revises and supercedes the staff positions stated in NUREG-0313 with the following principal differences:
The guidelines for reducing the intergranular stress corrosion cracking (IGSCC) susceptibility have been extended to cover ASME Code Class 2 and 3 piping.
Augmented inservice inspection requirements for nonconforming safe ends have been included.
The inservice inspection sampling schemes have been updated to comply with the most recent Code edition that has been subsequently accepted by the NRC.
The recommendations of NUREG-0531 (1978 Pipe Crack Study Group report), which cannot be implemented immediately without further NRC evaluation, have been identified.
The developmental items identified in this report are for future improvements and are not required for the present plant safety or for the resolution of Generic Task No. A-42.
The staff concludes that, pending implementation of the guidelines of this report, IGSCC in BWR pressure boundary piping, although undesirable, will not pose an undue risk to the health and safety of the public.
Leaks and cracks in the heat affected zones (HAZs) of welds that join austenitic stainless steel piping and associated components in BWRs have been observed since mid-1960.
Prior to September 1974, the affected (cracked) piping was mainly Type 304 stainless steel with diameters of 8 inches or less.
All the cracks were attributed to IGSCC due to the combination of high local stress, sensitization of material, and high oxygen content in the water.
In each case, it was believed that the problem had been corrected or substantially reduced by better control of welding, contaminants, and design. >
During the last quarter of 1974, a number of incidents of IGSCC in weld HAZ of 4-inch-diameter recirculation bypass lines and in 10-inch-diameter core spray lines were observed.
Following these occurrences, the Nuclear Regulatory Commission (NRC) formed a Pipe Crack Study Group (PCSG) in January 1975 to (a) investigate the cause, extent, and safety implications of cracks, (b) make an interim recommendation for operating plants, and (c) recommend corrective actions to be taken by future plants.
In October 1975, the Study. Group pub-lished its report, NUREG-75/067, " Technical Report, Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants" (Ref. 2).
During the same general time span, the General Electric Company (GE) conducted an independent evaluation of cracking problems and submitted their findings and recommendations to the NRC (NED0-21000,
" Investigation of Cause of Cracking in Austenitic Stainless Steel Pipes,"
(Raf. 3).
Following staff review of the recommendations made by the Study Group and GE, the staff issued an implementation document, NUREG-0313. This document, based on the information available at that time, sets forth the NRC technical positions consistent with the recommendations of the Study Group.
Since 1975, IGSCC has continued to be detected in recirculation bypass and core spray lines.
Incidence of IGSCC has also been observed in some stainless steel recirculation riser piping up to 12 inches in diameter in Japan and in large-diameter (>20 inches) recirculation piping in Germany.
These incidents, together with the questions concerning the reliability of ultrasonic inspec-tions, led to the formation of a new PCSG by NRC in September 1978.
The new Study Group was specifically chartered to address the following issues:
s The significance of the cracks discovered in large-diameter pipes relative to the conclusions and recommendations set forth in the referenced report (Ref. 2) and in its implementation document, I
NUREG-0313; Resolution of concerns raised over the ability of ultrasonic techniques to detect cracks in austenitic stainless steel; The significance of the cracks found in large-diameter sensitized safe ends, and any recor.cendations regarding the current NRC program for dealing with this matter; The potential for stress corrosion cracking in PWRs; and The significance of the safe end cracking at Duane Arnold relative to similar material and design aspects at other facilities.
In February 1979, the Study Group issued a report, NUREG-0531, " Investigation l
and Evaluation of Stress-Corrosion Cracking in Piping.,' Light Water Reactor Plants" (Ref. 4).
The principal conclusion of the Study Group is that IGSCC in large-diameter piping, while undesirable, will not be a hazard to public health and safety.
Nevertheless, the Study Group did recommend certain actions to be taken to improve plant reliability, reduce personnel exposure, and enhance.
m.
~,
the defense-in-depth approach.
The new Study Group reaffirmed that the conclusions and recommendations reported in NUREG-75/067 by the previous group and the implementing document, NUREG-0313, are still valid.
In addition, they discussed several additional ways to reduce the potential for IGSCC and also addressed IGSCC in safe ends.
During the same general time span, the General Electric Company conducted an independent evaluation of the recent cracking in large-diameter pipes and submitted their findings and recommendations to the NRC (Ref. 5).
The GE main conclusions are:
(a) IGSCC in Type 304 stainless steel weld HAZs remains to be a non-safety problem in spite of recent cracking in large-diameter pipes, and (b) GE approach outlined in NE00-21000 continues to be valid.
Following the issuance of the Study Group report in February 1979, the NRC noticed the availability of the report in the Federal Register and requested interested parties to provide any comments to the NRC by May 15, 1979.
A copy of that Notice is attached as Appendix 8.
Comments were requested so that the staff would have the benefit of industry and public comments prior to the development of its revised guidelines targeted for issuance in August 1979.
In response to the staff's request, comments from six organizations and individuals were received. These comments are summarized in Appendix C.
Again, following the issuance of the "For Comment" edition of NUREG-0313, Revision 1 in October 1979, the NRC noticed the availability of the report in the Federal Register on November 16, 1979, and requested interested parties to provide comments.
Comments from 11 organizations have been received.
All these comments were taken into consideration by the staff in developing its position as stated in this report.
The IGSCC occurs in a small percentage of the welds in BWR piping that contain.s relatively stagnant, intermittent or low-flow coolant.
Historically, these cracks have been discovered either by volumetric examination, by visual inspec-tion, or by leakage detection systems.
The growth pattern of the cracks is such that it is unlikely that these cracks would go undetected before they grow to significant size where the pipe function might be compromised.
- Further, because of the inherent high material toughness of austenitic stainless' steel piping, IGSCC is unlikely to cause a rapidly propagating failure resulting in a loss-of-coolant accident.
Although the likelihood is extemely low that these IGSCCs will propagate far enough to create a significant hazard to the public, the occurrence of such cracks is undesirable.
Measures should therefore be taken to minimize IGSCC in BWR piping systems to improve overall plant reliability, reduce personnel exposure associated with repair and inspection of degraded pipe, and enhance the defense-in-depth approach.
It is the purpose of this document to set forth the NRC staff's revised accept-able methods to reduce the IGSCC susceptibility of BWR piping and thus provide an increased level of reactor coolant pressure boundary and engineered safety features systems integrity.
Recognizing that complete compliance with these guidelines may not be practical, or even possible, for all plants, varying degrees of conformance to our guidelines are provided in Part II.
Corrosion-resistant materials for installation in BWR piping system, methods of testing, and processing techniques acceptable to the NRC are presented in Part III. -
For plants that cannot fully comply with the guidelines specified in Pa' t III r
of this document, varying degrees of augmented inservice inspection and leak detection requirements are established in Part IV.
The general recommendations that will lead to either limit the extent of IGSCC or improve the chance of detecting such IGSCC are outlined in Part V.
They are for future improvements and are not required for the present plant safety or for the resolution of Generic Task No. A-42.
II.
IMPLEMENTATION OF MATERIAL SELECTION, TESTING, AND PROCESSING GUIDELINES A.
For plants under review, but for which a construction permit has not been issued, all ASME Code Class 1, 2, and 3 lines should conform to the guidelines stated in Part III.
B.
For plants that have been issued a construction permit but not an operating license, all ASME Code Class 1, 2, and 3 lines should conform to the guidelines stated in Part III unless it can be demon-strated to the staff that implementing the guidelines of Part III would result in undue hardship.
For cases in which the guidelines of Part III are not complied with, additional measures should be taken for Class 1 and 2 lines in accordance with the guidelines stated in Part IV of this document.
C.
For plants that have been issued an operating license, NRC designated
" Service Sensitive" lines (Part IV. B) should be modified to conform to tne guidelines stated in Part III, to the extent practicable.
When " Service Sensitive" and other Class 1 and 2 lines do not meet the guidelines of Part III, additional measures should be taken in accordance with the guidelines stated in Part IV of this document.
Lines that experience cracking during service and require replacement should be replaced with piping that conforms to the guidelines stated in Part III.
III.
SUMMARY
OF ACCEPTABLE METHODS TO MINIMIZE CRACK SUSCEPTIBILITY - MATERIAL SELECTION, lESTING, AND PROCESSING GUIDELINES A.
Selection of Materials Only those materiale described in paragraphs 1 and 2 below are acceptable to the NRC for installation in BWR ASME Code Class 1, 2, and 3 pipirq systems.
Other materials may be used when evaluated and accepted by the NRC.
1.
Corrosion-Resistant Materials All pipe and fitting material including safe ends, thermal sleeves, and weld metal should be of a type and grade that has been demonstrated to be highly re istant to oxygen-assisted stress corrosion in the as-installed condition. Materials that have been so demonstrated include ferritic steels, " Nuclear "These materials have controlled low carbon (0.02% max) and nitrogen (0.1% max) contents and meet all requirements, including mechanical property requirements, of ASME specification for regular grades of Type 304 or 316 stainless steel pipe.
Grade" austenitic stainless steels,* Types 304L and 316L austent-tic stainless steels, Type CF-3 cast stainless steel, Types CF-8 and CF-8M cast austenitic stainless steel with at least 5%
ferrite, Type 308L stainless steel weld metal, and other austen-itic stainless steel weld metal with at least 5% ferrite content.
Unstabilized wrought austenitic stainless steel without controlled low carbon has not been so demonstrated except when the piping is in the solution-annealed condition.
The use of such material (i.e., regular grades of Types 304 and 316 stainless steels) should be avoided.
If such material is used, the as-installed piping including welds should be in the solution-annealed condition. Where regular grades of Types 304 and 316 are used and welding or heat treatment is required, special measures, i
such as those described in Part III.C, Processing of Materials, shald be taken to ensure that IGSCC will not occur.
Such measures may include (a) solution annealing subsequent to the welding or heat treatment, and (b) weld cladding of materials to be welded using procedures that have been demonstrated to reduce residual stresses and sensitization of surface materials.
2.
Corrosion-Resistant Safe Ends and Thermal Sleeves All unstabilized wrought austenitic stainless steel materials used for safe ends and thermal sleeves without controlled low carbon contents (L grades and Nuclear Grade) should be in the solution-annealed condition.
If as a consequence of fabrica-tion, welds joining these materials are not solution annealed, they should be made between cast (or weld overlaid) austenitic j
stainless steel surfaces (5% minimum ferrite) or other materials having high resistance to oxygen-assisted stress corrosion. The joint design must be such that any high-stress areas in unstabi-l lized wrought austenitic stainless steel without controlled low carbon content, which may become sensitized as a result of the welding process, is not exposed to the reactor coolant.
Thermal sleeve attachments that are welded to the pressure boundary and form crevices where impurities may accumulate should not be exposed to a BWR coolant environment.
l 8.
Testing of Materials l
For new installation, tests should he made on all regular grade l
stainless steels to be used in the ASME Code Class 1, 2, and 3 piping systems to demonstrate that the material was properly annealed and is not susceptible to IGSCC.
Tests that have been used to determine the susceptibility of IGSCC include Practices A* and E** of ASTM A-262, " Recommended Practices for
" Practice A - Oxalic acid etch test for classification of etch structures of stainless steels.
- Practice E - Copper-copper sulfate-sulfuric acid test for detecting susceptibiltiy to intergranular attack in stainless steels.
l l
Detecting Susceptibility to Intergranular Attack in Stainless Steels" and the electrochemical potentiokinetic reactivation (EPR) test.
The EPR test is not yet accepted by the NRC.
It the EPR test is 1
used, the acceptance criteria applied must be evaluated and accepted i
by the NRC on a case-by-case basis.
/
C.
Processing of Materials Corrosion-resistant cladding with a duplex microstructure (5% minimum ferrite) may be applied to the ends of Type 304 or 316 stainless steel pipe for the purpose of avoiding IGSCC at weldments.
Such cladding, which is intended to (a) minimize the HAZ en the pipe inner surface, (b) move the HAZ away from the highly stressed region next to the attachment weld, and (c) isolate the weldment from the environment, may be applied under the following conditions:
1.
For initial construction, provided that all of the piping is solution annealed after cladding.
2.
For repair weldir:. and modification to in place systems in operating plants and plants under construction.
When the repair welding or modification requires replacement of pipe, the replacernent pipe should be solution-annealed after cladding.
Corrosion-resistant cladding applied in the " field" (i.e.,
without subsequent solution annealing of the pipe) is acceptable only on that portion of the pipe that has not been removed from the piping system.
Other " field" applications of corrosion-resistant cladding are not acceptable.
Oth processes that have been found by laboratory tests to minimize stresses and IGSCC in austenitic stainless steel weldments include induction heating stress improvement (IHSI) and heat sink welding (HSW).
Although the use of these processes as an alternate to augmented inservice inspection is not yei, accepted by the NRC, these processes may be permissible and will be considered on a case-by-case basis provided acceptable supportive data are submitted to the NRC.
IV.
INSERVICE INSPECTION AND LEAK DETECTION REQUIREMENTS FOR BWRs WITH VARYING DEGREES OF CONFORMANCE TO MATERIAL SELECTION, TESTING, AND PROCESSING GUIDELINES A.
For plants whose ASME Code Class 1, 2 and 3 pressure boundary piping meets the guidelines of Part III, no augmented inservice inspection or leak detection requirements beyond those specified in the 10 CFR 50.55a(g), " Inservice Inspection Requirements" and plant Technical Specifications for leakage detection are necessary.
B.
ASME Code Class 1 and 2 pressure boundary piping that does not meet guidelines of Part III is designated " Nonconforming" and must have additional inservice inspection and more stringent leak detection requirements.
The degree of augmented inservice inspection of such piping depends on whether the specific " Nonconforming" piping runs are classified as " Service Sensitive." The " Service Sensitive" lines were and will be designated by the NRC and are defined as those that have experienced cracking of a generic nature, or that are considered to be particularly susceptible to cracking because of a combination of high local stress, material condition, and high oxygen content in the relatively stagnant, intermittent, or low-flow coolant.
Currently, for the nonconforming ASME Code Class 3 piping, no additional inservice inspection beyond the Section XI visual examination is required.
Examples of piping considered to be " Service Sensitive" include but are not limited to:
core spray lines, recirculation riser lines,*
recirculation bypars lines (or pipe extensions / stub tubes on plants where the bypass lines have been removed), control rod drive (CRD) hydraulic return lines, isolation condenser lines, recirculation inlet lines at safe ends where crevices are formed by the welded thermal sleeve attachments, and shutdown heat exchanger lines.
If cracking should later be found in a particular pjping run and considered to be generic, it will be designated by the NRC as
" Service Sensitive."
Leakage detection and augmented inservice inspection requirements for " Nonconforming" lines and " Nonconforming, Service Sensitive" lines are specified below:
1.
" Nonconforming" Lines That Are Not " Service Sensitive" a.
Leak Detection:
The reactor coolant leakage detection systems should be operated under the Technical Specification t
requirements to enhance the discovery of unidentified leakage that may include through-wall cracks developed in austenitic stainless steel piping.
(1) The leakage detection system provided should include sufficiently diverse leak detection methods with adequate sensitivity to detect and measure small leaks in a timely manner and to identify the leakage sources within the practical limits.
Acceptable leakage detection and monitoring systems are described in Section C, Regulatory Position of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."
- Since no IGSCC has been observed in the domestic plants and in view of the possible high radiation exposure to the inspection personnel, surveillance i
and monitoring means other than those specified in Section IV of this report for recirculation riser lines will be considered on a case-by-case basis. _ -
Particular attention should be given to upgrading and calibrating those leak detection systems that will provide prompt indication of an increase in leakage rate.
Other equivalent leakage detection and collection systems will be reviewed on a case-by-case basis.
(2) Plant shutdown should be initiated for inspection and corrective action when any leakage detection system indicates, within a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less, an increase in rate of unidentified leakage in excess of 2 gallons per minute or its equivalent, or when the total unidentified leakage attains a rate of 5 gallons per minute or its equivalent, whichever occurs first.
For sump level monitoring systems with l
fixed-measurement interval method, the level should be monitored at 4-hour intervals or less.
(3) Unidentified leakage should include all leakage other than:
(a) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered, and conducted to a sump or collecting tank, or (b) Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the op'erations of unidentified leakage monitoring systems or not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.
b.
Augmented Inservice Inspection:
Inservice inspection of the " Nonconforming, Nonservice Sensitive" lines should be conducted in accordance with the following program:*
(1) For ASME Code Class 1 components and piping, each 0
pressure-retaining dissimilar metal weld subjact te inservice inspection requirements of Section XI should be examined at least once in no more than 80 months (two-thirds of the time prescribed in the ASME Boiler and Pressure Vessel Code Section XI).
Such examination should include all internal attachment welds that are not through-wall welds but are welded to or form part of the pressure boundary.
- This program is largely taken from the requirements of ASME Boiler & Pressure Vessel Code,Section XI, referenced in the paragraph (b) of 10 CFR 50.55a,
" Codes and Standards." __
(2) The following ASME Code Class 1 pipe welds subject to inservice inspection requirements of Section XI shotld be examined at least once in no more than 80 months:
(a) All welt's at terminal ends
- of pipe at vessel nozzles; (b) All welds having a design combined primary plus secondary stress range of 2.4S, or more; (c) All welds having a design cumulative fatigue usage factor of 0.4 or more; and (d) Sufficient additional welds with high potential for cracking to make the total equal to 25% of the welds in each piping system.
(3) The following ASME Code Class 2 pipe welds, subject to inservice inspection requirements of Section XI, in each residual heat removal systems, emergency core cooling systems, and containment heat removal systems should be examined at least once in no more than 80 months:
(a) All welds of the terminal ends of pipe at vessel nozzles, and (b) At least 10% of the welds selected proportionately from the following categories:
(i) Circumferential welds at locations where the stresses under the loadings resulting l
from am plant conditions as calculated by the sum of Equations (9) and (10) in NC-3652 exceed 0.8 (1.2Sh + S )I A
(ii) Welds at terminal ends of piping, including branch runs; (iii) Dissimilar metal welds; (iv) Welds at structural discontinuities; and (v) Welds that cannot be pressure tested in accordance with IWC-5000.
" Terminal ends are the extremities c' piping runs that connect to structures, components (such as vessels, pumps, valves)~or pipe anchors, each of which acts as rigid restraints or provides at least two degrees of restraint to piping thermal expansion.
9_
The welds to be examined shall be distributed approximately equally among runs (or portions of runs) that are essentially similar in design, size, system function, and service conditions.
(4) The following ASME Code Class 2 pipe welds in systems other than residual heat removal systems, emergency core cooling systems, and containment heat removal systems, which are subject to inservice inspection requirements of Section XI, should be inspected at least once in no more than 80 months:
(a) All welds at locations where the stresses under the loadings resulting from " Normal" and " Upset" plant conditions including the operating basis earthquake (OBE) as calculated by the sum of Equations (9) and (10) in NC-3652 exceed 0.8 (1.2Sh + S )I A
(b) All welds at terminal ends of piping, including branch runs; (c) All dissimilar metal welds; (d) Additional welds with high potential for cracking at structural discontinuities
- such that the total number of welds selected for examination equal to 25% of the circumferential welds in each piping system.
(5) If examination of (1), (2), (3), and (4) above conducted during the first 80 months reveal no incidence of stress corrosion cracking, the examination frequency thereafter can revert to 120 months as prescribed in Section XI of the ASME Boiler and Pressure Vessel Code.
(6) Sampling plans other than those described in (2),
(3), and (4) above will be reviewed on a case-by-case basis.
2.
" Nonconforming" Lines That Are " Service Sensitive" a.
Leak Detection:
The leakage detection requirements, described in IV.B.l.a above, should be implemented.
" Structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc., conforming to ANSI Standard B 16.9) and pipe branch connections and fittings..
i b.
Augmented Inservice Inspection:
(1) The welds and adjoining areas of bypass piping of the discharge valves in the main recirculation loops, and of the austenitic stainless steel reactor core spray piping up to and including the second isolation valve, should be examined at each reactor refueling outage or at other scheduled plant outages.
Successive examination need not be closer than 6 months, if outages occur more frequently than 6 months.
This requirement applies to all welds in all bypass lines whether the 4-inch valve is kept open or closed during operation.
In the event these examinations find the piping free of unacceptable indications for three successive inspections, the examination may be extended to each 36-month period (plus or minus by as much as 12 months) coincident with a refueling outage.
In these cases, the successive examination may be limited to all welds in one bypass pipe run and one reactor core spray piping run.
If unacceptable flaw indications are detected, the remaining piping runs in each group should be examined.
In the event these 36 month period examinations reveal no unacceptable indications for three success-ive inspections, the welds and adjoining areas of these piping runs should be examined as described in IV.B.l.b(1) for dissimilar metal welds and in IV.B.l.b(2) for other welds.
(2) The dissimilar metal welds and adjoining areas of other ASME Code Class 1 " Service " Sensitive" piping should be examined at each reactor refueling outage or at other scheduled plant outages.
Successive examinations need not be closer than 6 months, if outages occur more frequently than 6 months.
Such examination should include all internal attachments that are not through-wall welds but are welded to or form part of the pressure boundary.
(3) The welds and adjoining areas of other ASME Code Class 1 " Service Sensitive" piping should be examined using the sampling plan described in IV.B.1.b(2) except that the frequency of such examinations should be at each reactor refueling outage or at other scheduled plant outages.
Successive examinations need not be closer than 6 months, if outages occur more frequently than 6 months. _ _ -
(4) The adjoining areas of internal attachment welds in recirculation inlet lines at safe ends where crevices
)
are formed by the welded thermal sleeve attachments should be examined at each reactor refueling outage or at other scheduled plant outages.
Successive examinations need not be closer than 6 months, if outages occur more frequently than 6 months.
(5)
In the event the examinations described in (2), (3) and (4) above find the piping free of unacceptable indications for three successive inspections, the examination may be extended to each 36-month period (plus or minus by as much as 12 months) coinciding with a refueling outage.
In the event these 36-month period examinations reveal no unacceptable indications for three successive inspections, the frequency of examination may revert to 80-month periods (two-thirds the time prescribed in the ASME Code Section XI).
(6) The area, extent, and frequency of examination of the
)
augmented inservice inspection for ASME Code Class 2
" Service Sensitive" lines will be determined on a case-by-case basis.
j 3.
Nondestructive Examination (NDE) Requirements The method of examination and volume of material to be examined, the allowable indication standards, and examination procedures should comply with the requirements set forth in the applicable Edition and Addenda of the ASME Code,Section XI, specified in paragraph (g), " Inservice Inspection Requirements," of 10 CFR 50.55a, " Codes and Standards."
In some cases, the code examination procedures may not be effective for detecting or evaluating IGSCC and other ultrasonic (UT) procedures or advanced nondestructive examination techniques may be required to detect and evaluate stress corrosion cracking in austenitic stainless steel piping.
Improved UT procedures have been developed by certain organizations.
These improved UT detection and evaluation procedures that have been or can be demonstrated to the NRC to be effective in detecting IGSCC should I
be used in the inservice inspection.
Recommendations for the development and eventual implementation of these improved techniques are included in Part V.
V.
GENERAL RECOMMENDATIONS The measures outlined in Part III of this document provide for positive actions i
that are consistent with current technology.
The implementation of these actions should markedly reduce the susceptibility of stainless steel piping to stress corrosion cracking in BWRs.
It is recognized that additional means could be used to limit the extent of stress corrosion cracking of BWR pressure boundary piping materials and to improve the overall system integrity.
These include plant design and operational procedure considerations to reduce system exposure to potentially aggressive environment, improved material selection, special fabrication and welding techniques, and provisions for volumetric inspection capability in the design of weld joints.
The use of such means to limit IGSCC or to improve plant system integrity will be reviewed on a case-by-case basis.
Although the items identified below are not required for the present plant safety, they may be expected to lead to means of limiting the extent of IGSCC and improving the chance of detecting such IGSCC.
Some of these items have not yet been fully developed (or have recently been developed) and have not yet been accepted by the NRC.
The acceptance of these items by the NRC may permit the reduction of augmented inservice inspection.
Specifically, areas recommended in NUREG-0531 that require further consideration are as follows:
A.
Improved ultrasonic inspection methods.
Such methods should be included in the ASME Code or included in a Regulatory Guide.
B.
Development and implementation of an improved focused inservice inspection program based on stress rule index, material of con-struction, history of cracking, etc.
C.
Improved weld joint design to ensure that required examinations can be performed effectively.
D.
Reduction of oxygen content in reactor coolant during all phases of reactor operation by water chemistry control, de-aeration of systems, etc.
E.
Minimization of stagnant or low-flow coolant pressure boundary piping.
F.
Evaluation of newly developed alternate corrosion-resistant materials in BWR environment.
G.
Evaluation of improvement of material corrosion resistance by alternate methods such as heat sink welding, induction heating stress improvement, etc.
H.
Evaluation of the electrochemical potentiokinetic reactivation technique for detecting and quantifying the degree of sensitization in stainless steel piping.
I.
Continued evaluation and verification of leak before break concept.
J.
Evaluation and implementation of leakage detection capability to improve early detection of small leaks. __
VI.
REFERENCES 1.
U.S. Nuclear Regulatory Commission, " Technical Report on Material
' Selection and Processing Guidelines for BWR Cooltat Pressure Boundary Piping," USNRC Report NUREG-0313, July 1977.*
2.
U.S. Nuclear Regulatory Commisison, " Technical Report, Investigation and Evaluation of cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants," USNRC Report NUREG-75/067, October 1975.*
3.
General Electric Company, "Investiaation of cause of cracking in Austenitic Stainless Steel Pipes," NEDC-21000, July 1975.**
4.
U. S. Nuclear Regulatory Commission, " Investigation and Evaluation of Stress-Concession cracking in Piping of Light Water Reactor Plants,"
USNRC Report NUREG-0531, February 1979.*
5.
Letter from G. Sherwood, GE, to V. Stello, NRC,
Subject:
General Electric Meeting with NRC on IGSCC, dated September 12, 1978.**
- Available for purchase from NRC/690 Sales Program, Division of Technical Information and Document Control, U. S. Nuclear Regulatory Commision, Washington, DC 20555 and National Technical Information Service, Springfield, Virginia 22161.
- Available in NRC Public Document Room for inspection and copying for a fee.
APPENDIX A TASK A-42 PIPE CRACKS IN BOILING WATER REACTORS 1
Division of Operating Reactors (DOR)
Lead Supervisor:
L. C. Shao, Acting Assistant Directar for Engineering Programs Task lianager:
C. Y. Cheng, DOR:EB Applicability:
General Electric Boiling Water Reactors Projected Completion Date:
December 1979 1
i s
1 A-1
1.
DESCRIPTION OF PROBLEM Leaks and cracks in the heat-affected zones (HAZs) of welds that join austenitic stainless steel piping and associated components in BWRs have been observed since mid-1960s.
Prior to September 1974, all affected piping was Type 304 stainless steel with diameters of eight inches or less.
All the cracks were attributed to intergranular stress corrosion cracking (IGSCC) due to the combination of high local stress, sensitization of material, and high oxygen content in the water.
During the last quarter of 1974, a number of incidents of IGSCC in weld HAZs of i
4-inch diameter recirculation bypass lines and in 10-inch diameter core spray lines were again observed.
Following these occurrences, the NRC formed a Pipe Cracking Study Group (PCSG) to (a) investigate the cause of cracks, (b) make an interim recommendation for operating plants, and (c) recomend corrective actions to be taken by future plants.
The study Group published its report a
(NUREG-75/067) in October 1975 which contains several recommendations to reduce the incidence of IGSCC in sensitized stainless steel piping.
Following staff review of the Study Group's recommendations, the staff issued an inplementation document (NUREG-0313) which established staff positions consistent with the recomendations of the Study Group.
The staff has been in the process of implementing these positions over the last couple of years for operating plants and for plants under review for an operating license.
Since 1975 IGSCC has continued to be found in recirculation bypass and core spray lines.
Incidentsof IGSCC have also been observed in some stainless steel recirculation riser piping up to twelve inches in diameter and in large diameter (>20 inches) recirculation pipinr; in foreign countries.
Cracks in these large recirculation lines had not teer observed prior to 1975.
These incidents, together with the reported questions concerning the reliability of ultrasonic inspections (UT), led to the activation of a new PCSG by NRC in September 1978.
The new Study Group was specifically chartered to reexamine the conclusions and recommendations of the 1975 PCSG report in view of cracks recently discovered in large diameter pipes.
Particular attention was given to the significance of cracking found in large recirculation lines, to evaluate the capability of nondestructive examination (NDE) methods to detect IGSCC and, in addition, i
to assess the significance of the safe-end cracking at Duane Arnold relative to similar material and design aspects at other facilities.
The 1978 Study Group completed its evaluation and published the NUREG-0531 report in February 1979.
The most important finding of this investigation I
was that the conclusions and recontendations reached in NUREG-75/067 by the previous PCSG and the implementation document, NUREG-0313, are still valid.
The present Study Group not only reaffirmed the conclusions and recomendations reached by the previous group but also presented some new ideas to reduce the potential for IGSCC based on the operating experience since 1975 and the recent pipe cracking in large diameter pipes.
In addition, the present Study Group has addressed IGSCC in safe-ends and has reached conclusions and recommendations concerning them which were not discussed by the previous Study Group.
Because of these new ideas and issues addressed by the 1978 PCSG, the implementation 4
document NUREG-0313 needs to be updated to incorporate the latest recomendations made by the present Study Group.
A-2
2.
PLAN FOR PROBLEM RESOLUTION A.
Approach The problem will be resolved by identifying the new conclusions and re-comendations reached by the present PCSG by carefully studying and com-paring the conclusions and recomendations made in NUREG-75/067, NUREG-0313, and NUREG-0531.
The implementation document NUREG-0313 will then be revised to incorporate those new recomendations which can be implemented imediately.
For those new recomendations which will require further study before it can be implemented, a plan for establishing the staff l
position on each recommendation will be proposed.
B.
End Product The end product of this activity will be a NUREG report documenting the updated staff position on material selection and processing guidelines for BWR piping based on recommendations made by the present PCSG.
This report will be issued approximately in Mid-August 1979.
C.
Tasks C-1.
Revision of NUREG-0313 " Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" Review and identification of those new conclusions and recommendations in NURES-0531 which can be implemented imediately.
The specific effort will include updating the implementation document NUREG-0313 to incorporate these new recommendations. This subtask will be accomplished in Mid-August 1979.
l C-2.
Staff Recomendation of Follow-on Efforts to Reduce the Potential for l
IGSCC in BWR Piping Those conclusions and recomendations of NUREG-0531 which would require further study before the staff position can be established will be identified.
In addition, a plan for establishing such a position will be recommended.
This subtask will also be completed approximately in Mid-August 1979.
However, the technical activities for these follow-on efforts will definitely not be completed within the time span specified for this activity.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION OF TASK For new plants or plants under construction and operating plants, we have conclud-ed that, pending completion of this task, continued plant operatio.: and licensing do not constitute an undue risk to the health and safety of the public for the following reasons:
Although the augmented inservice inspection programs required by NRC cannot detect all IGSCC, it has demonstrated to be effective in locating most instances of IGSCC prior to cracks propagating through the wall.
A-3
The leak detection system employed as a monitoring system has been effective in alerting the plant operators of primary system leakage that could result from a through-wall crack.
.%dden failure or significant loss-of-coolant is not expected from arough-wall cracks prior to a period of leakage.
Should a large through-wall crack develop, go undetected by NDE inspections, and by continuous leak detection devices, and subse-quently should a rupture of the line occur causing a loss-of-coolant accident, the design of a nuclear power plant is such that protection is still provided for the public health and safety.
To sumarize, the various NRC actions taken to date ensure that IGSCC does not pose an immediate safety problem to operating plants and thus constitute an acceptable basis for continut.d plant operation ar,d licensing.
4.
NRC TECHNICAL ORGANIZATION INVOLVED A.
Engineering Branch (EB), Division of Operating Reactors, has the oyerall lead responsibility to see this TAP to its completion.
This includes reyjew and evaluation of the subject NUREG reports to establish the implementation guidelines with particular emphasis on operating plants, and final issuance of a NUREG report.
In addition, EB will have the lead responsibility of identifying long-term follow-on efforts and recomending plans for establish-ing the implementing guidelines for thise issues.
Manpower Estimates: 4 man-months FY 1979 8.
Materials Engineering Branch (MTEB), Division of Systems Safety, has the lead responsibility of establishing the implementation guidelines for new plants and plants under construction.
MTEB will have direct input to the revision of NUREG-0313.
MTEB will also identify long-term follow-on efforts and recomend plans for establishing staff position on these issues.
Manpower Estimates:
3.5 Man-Months FY 1979 TECHNICAL ASSISTANCE No technical assistance is needed for the present tasks.
However, technical assistance may be required for the identified follow-on efforts.
l 6.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES i
No assistance from other NRC offices is required for Subtasks C-1 and C-2.
However, some assistance may be needed for the follow-on efforts identified under l
l Subtask C-2.
All research and developmental programs aiming to increase or maintain the integrity of BWRs piping will definitely assist us in establishing l
the implementation guidelines for the follow-on efforts.
Specifically, A.
Office of Standards Development l
l Structures and Components Standards Branch / DES is currently funding EG&G to l
develop a Regulatory Guide on "UT of Austenitic Stainless Steel Piping."
A-4 l
This guide will provide a UT performance standard or procedure which will significantly increase the detection capability for IGSCC in austenitic stainless steel piping.
B.
Office of Nuclear Regulatory Research Metallurgy and Materials Branch /RES is currently fundir.g the Pacific Northwest Laboratories to study the " Reliability of Ncn-destructive Examination" aimed to pinpoint the strengths and weaknesses of NDE and recommend the appropriate experimental programs to increase the reliability of flaw detec-tion.
7.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS No major interactions with outside organizations are anticipated for the subtasks.
However, an extensive interaction with outside organizations will be necessary for the follow-on efforts.
This interaction involves information exchanges with licensee, GE, industry research institutes, and national labs that are active in research and development of methos to reduce the potential for IGSCC or to detect the occurrence of IGSCC. An information exchange with foreign regulatory and inspection organizctions is also expected.
8.
POTENTIAL PROBLEMS No difficulties have been anticipated in achieving this task.
However, some delay in achieving the follow-on efforts, if the task is expanded, might be expected because of the long-term nature of the problem and the necessary extensive interactions with other organizations.
l l
A-5
NOTICES 14655 APPENDIX B NRC NOTICE IN THE FEDERAL REGISTER REQUESTING PUBLIC COMMENT ON NUREG - 0531
[7590-01-M]
00MtsTic LICENilNG OF rRODUCT10N AND UTitif.AT10N FAOUTits in... ;,.is.a. d s..h
.e si.
C.,,.o.a C,..uas la r;rias of tishi wei.e
(
a..o., rs.ne.
l AGENCY: U.S. Nuclear Regulatory Commission.
ACTION: Request for public comment l
on NUREG-0531 Investigation and Evaluation of Stress Corrosion Crack ing in Piping of IJaht Water Reactor Plants" February 1979.
SUMMARY
- On September 14,19*t8, the Nuclear Regulatory Commission
, established a new Pipe Crack Study Group. The Group was to evaluate recent pipe and safe end cracking ex.
perience relative to previous staff con.
, clusions and recommendations. The
' NRC seeks public comment en the report which summarizes the Group's
- review and conclusions.
DATES: The public ecwnent period expires May 15.1979.
FOR FURTHER INFORMATION CONTACT:
Darrell G. Eisenhut, Deputy Direc.
tor for Operating Reactors. Division FtOtt At tlGISTER, VOL 44, NO. 50 -TutsDAY, MARCH 13,1979 B-1
14555 NOTICES cf Operating Reactcrs. Offlee cf Nu-e Pix Ccn!!ruratfen a.d Stress citar Reatter Regt:stlen. U.S. Nu.
14ve:.s clear Regula. cry Ccmmission. Wash.
e Duane Arncid Safe-End Cracki.~.g br.on. D C. 2:535. (Phene: 3C1-492-e Methods cf Detecting Cracks 7:21) e Significance cf Cracks SUPP EMENTARY INPORMATION:
o Rmnt Dmicpment Re:eM 2 Cen*rc! and Detecuen cf MC In 1975, a P2pe Cracking Study Group m n m r cf 1 ese te;u !n the sas establishad by the Un!ted States centext cf changes cecurrbs since the yuclear Reguintery Ccmmissica pre;arauca cf NUREG-iS/C67 led to (USNRC) to rester intergranu:ar the preparati n cf specific conc! Wens stress-ectros ca crack 5g (IGSCC) in and recer.=endat!cas re etant L the Bct!!ng Water Reacters (BWRsk The current status cf IGSCC, the s!rn!!!-
Crcup reperied its findings cencern-cance cf tne ;rctlem, and the re!:stil-irg stress.ccrresten cracking in by pass !!y cf deteeden and measures avanat:e knes and core 3 ray ;t;!ng cf austen-to ccrrect er min!=1:e IGSCC in exist-tit sta*.nless steel in a repert. TechM-Lag and future ; ants These cenclu-ce! Repcsintesticcticn cnd Esthe-signs nn: greeze,ngs::g._, gre gre.
Ifew c/ Creckin; in As.ste-sitie S: cts
- sented in the nesly issued PCSG less Stre! P:;t*.c cf 3: sling Wcter Re-g, pe r*,,
ceter Ilcs!.s (NUR EG-75/Ctit ne NRC staff vill rerfer the Study During 1978. IGSCC ru rep:rted G cup repcrt and its ccnclu.s!cns/ rec.
fcr the nrst tine in large diameter cmmendauens and the ;ut!!: ec=-
- !;ing in a BWR. "This d!.tectery. to-ments recetted during this ec=. ment Stther with cuest!cns ccactining the period. Pcuestr th!.s review. the staff et;atD!ty cf ultru
- nic detect!cn stu decide what further actions, if
- ethets to detect s: an cracks led to any, are required for t.be U:ensing and the formation cf a new Pi:e Crack c;eration cf reacters.
Study Grcup (PCSG) by UShAC cn Re uesta fer a single cc 7 cf the Se:temte* 14.19*8.
rexrt should te :: ace in sT!!!ng to The cht.rter cf the f.er PCSG sas to U.S. Nuclear R~egulatcry Cc==ssicn.
3:eciftta. :y address the five foUcsts Wuhi.r.cn. D.C. :C155. Attenti:n Di-quest te ns-recter. D!rtsien cf Tec.6.nl:a3 I..fer=a.
- 1. The s!gninta ce cf the cracks t!:n and Decument CenticL d Sect eredi in larse<*iameter ;1 pes rel.
C:m=ents en this repcrt shou:2 te attre to the conclusiens and reecm. sent to the Office cf Nu:3 ear Rea:ter
!=endations set fcrth in the referenced Reg.:lation. U.S. Nuclear Regulato y repcrt (NUREG-75/067) and its t= le.
Cc=.misa!:n. Washingten. D.C. 20555, mentatien doeur ent. NUREG-C313; Attentle.: Du uty Dire:ter. Divisica
- 2. Res:Iutien cf the concerns raised cf Opera;5g Rea: ten. ne ecm:nent cser the sbi!!!y to u.se ultrasenic tech-pened es; ires May 15,1971. Cc;tes cf niques to detect cracks in austentti: an ecmments it:etred rtII be art!!atte stain:ess steet:
fer examinatica in the Cem. !ssien's
- 3. The sign!ficante cf cra:ks fcund Puth: Document Re:n. 1717 H In large. diameter sensitized safe ends Street N.W. Washingten. D.C.
and any reecm.endat!:n.s regarding Dated at Bethesda Md., this 6th day the cur tnt NRC preg arn for dealing cf March,Ig;p.
uth this reatter[-
at Regdam Cem-
- 4. The pctent a1 for stress corrosien cracking in PWRs-
- 5. Damine the significance of cra:k.
VI:rca S;1:.*c. Jr.,
ing in the Inccnel safe ends that has Dfreeter. Otrision c/ Operctist been esperienced at the Duane Arnc1d Recciars. Office c/Nuciecr Re.
Oxrating Patluty, and develop any actor Resulsfios recomr=endations regarding NRC ne-tm Doc.M-n:S 7".)ed 312-;st a.as amm3 tiens ta. ken or to be taken?
The PCSG limited the secpe of the study to BWR and PWR ;! ping and sa.fe ends attached to the reactor pres-sure sessel The PCSG reviewed ex.!st-mg infer nation-cither that rentained
.: wrttien reecrds c,r that ecile ted th cuch meetings in this ecuntry and
.. fcre:gn ecuntnes. The specthe areas tensidered are presented in the chap-ters ci th3 re;crt:
eBWR Crac king D;erience and Cerreem e Aettoa.s ePWR Craceng Dpertence and 0: reem e Acti:ns e MeTW Assn ated s!th P'.pe
- en-e Ret:ter Cc:: ant Che--is!*y Fl*(t at elCIsttt. WOL 44, h3. $!-. ?utitaY. Mat 041s. H?9 B-2
APPENDIX C
SUMMARY
OF PUBLIC COMMENTS ON NUREG-0531 In response to NRC's request, comments on NUREG-0531, " Investigation and Evalua-tion of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants" from the following six organizations and individuals were received.
Their substantive comments are summarized below:
A.
General Electric Company 1.
The use of regular grades of Type 304 and 316 stainless steel in BWR piping systems should be avoided unless carbon centent is restricted to 0.35% or less.
If regular grades without special carbon restric-tions are used, steps should be taken to ensure that IGSCC cannot occur.
Such measures may include non-welded applications, solution annealing, weld cladding, or other measures that have been adequately 1
tested to provide reasonable aasurance of reliable performance during the life of the plant.
(paragraph 4.8) 2.
The use of IHSI on existing plant welds raises some areas for futher investigation.
The effect of the treatment on existing cracks should be determined, as well as any effets of the thermomechanical IHSI cycle.
C-1
A 2
It is recognized that IHSI provides residual stress reduction in piping of all diameters.
GE-EPRI has planned tests to determine the extent of benefit to be derived from IH5I.
(paragraph 10.5.1) 3.
It is recommended that the recommendations contained in NUREG-0313 continue to be considered for operating plants and plants under review for an operating license or construction permit.
On a case by case basis plans should be developed for in-service inspection which.;ould improve the probability of early crack identification.
These plans should consider differences in stress, carbon content, degree of material sensitization and the frequency of past cracking incidents in other plants as well as other factors related to plant operation and inspection history.
(paragraph 2.11)
J 4.
Based on the incidence of IGSCC in recirculation-riser piping in the offshore plants, it is recommended that an augmented in-service inspection program considering the above factors be developed for i
these lines.
(paragraph 2.11) 5.
Further clarification is requested on the second recommendation relative to safe ends on reactor pressure vessels.
General Electric considers that special inspections of uncreviced safe-ends with tuning fork designs are not warranted.
(p. 7.4) 6.
To ensure that General Electric is aware of the complete list of NRC icentified field cracking incidents.in piping, it is requested that C-2 t
a detailed list be provided of these incidents by plant and line type.
(p. 2.1)
B.
Washington Public Power Supply System (WPPSS) 1.
WPPSS questions whether there is sufficient experience to warrant placing the riser lines in the service-sensitive category.
It only requires one minor extension of this logic to place the whole system in this category.
(paragraph 2.11) 2.
Fabrication of Materials-WPPSS feels that more discussion is warranted on the merits and adequacy of ASTM A-262 for acceptance of materials used in environments conducive to stress corrosion cracking.
By using the techniques in ASTM A-262, are we possibly accepting material which is partially sensitized prior to welding? (paragraph 4.2.3)
C.
Combustion Engineering - Power Systems j
l 1.
There appears to be an error in the specification for Boron concen-trations in Table 5.2, " Summary of PWR Reactor Coolant Chemistry Specifications".
The correct refueling boron concentration should be < 4400 ppm.
9 C-3
D.
Carolina Power and Light Company 1.
There is not sufficient justification for reclassifying the recir-culation - riser piping as nonconforming, service sensitive line.
(Recommendation 2.11.1) 2.
It is not practical to require utilities to reclassify their welded attachments as nonconforming, service sensitive lines.
The welds, in most cases, do not have configurations that will allow ultrasonic inspections.
(Recomraendation 7.4.1)
E.
Lawrence Livermore Laboratory 1.
Long term effects of redistribution of stress must be considered when a'ny heating and cooling cycle is superimposed on an existing welding process.
There is the possibility that cracks that occur could propagate without arrest because of the new stress distribution created by IHSI.
(p. 10.4) 2.
The report does not state how the results of A262 A and E compare with the lots of stainless steel which have experienced IGSCC in the BWR environment. 'There should be more discussion of electrochemical potentiokirratic reactivation technique (p. 4.3).
3.
Since the critical level of sensitization is probably a critical level of chromium depletion around the carbides, measuresments which C-4
emphasize these critical parameters should be the basis of regulatory requirements.
(p. 4.4) 4.
Further study is needed on the role of residual stress distribution on crack growth.
The residual stress distribution in circumferential welds tends to promote cracking all around the circumference.
As the crack extends around a significant portion of the inside wall, the residual stresses in the axial direction should increase and ac. elerate the crack growth.
(p. 6.4) 5.
The tearing modulus concept appears to be still at a research level.
A major comprehensive study of this topic appears to be justified.
- p. 9.1) 6.
The relation-of leakage rate to crack size should be studied relative to its usage as a criterion for crack detection.
(p. 8.5) l 7.
The 3-D presentation of internal defects by acoustical holography certainly warrants consideration as a complementary technique to ultrasonic testing.
F.
Southwest Research Institute (SWRI) 1.
SWRI agrees with Conclusions 1, 2, 3, and 4 of Chapter 8.
- However, it should be noted that sizing is not as important as the detecticn C-5
of IGSCC.
Once a reflector is identified as IGSCC, the area must be repaired under present requirements.
Therefore, there is no need to size the cracts at this time.
(p. 8.7) 2.
It is not SWRI's opinion that all reflectors observed within the HAZ should be classified as IGSCC.
However, all crack-like indications j
t within the HAZ of suspect austenitic welds should be classified as I
(p. 8.7) 3.
The technique mentioned in Recommendation 2 of Chapter 8 may serve to reduce radiation exposure.
However, it does nothing to improve the technical adequacy or credibility of the examination.
In fact, this approach may, at times, reduce the adequacy of the examination.
It should be noted that automatic recording and analysis of signal I
response and positional data for manual exami1ations will, in the l
1 near future, provide improved examinations wh le reducing radiation exposure to examination personnel.
(p.'8.7) l l
C-6
f,RCy oRu 335 U.S. NUCLE An REGULATORY COMMIS$10N BIBLIOGRAPHIC DATA SHEET NUREG-0313 Rev.1 4 TlTLE AND SUBT51LE (Add Volume No., o!apprminate)
- 2. (Leave blank)
Technical Report on Material Selection and Processing
. RECIPIENT'S ACCESSION NO.
Guidelines for BWR Coolant Pressure Boundary Piping
- 7. AUTHORIS)
- 5. DATE REPORT COMPLETED C.Y. Cheng, R.M. Gamble, A. Taboada, B. Turovlin July I"^"1980 k PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include 20 Codel
^
DATE REPORT ISSUED MONTH l YEAR U.S. Nuclear Regulatory Comnission July 1980 Office of Nuclear Reactor Regulation 3
ft,,,, y,,,,,
Washington, D.C.
20555
- 8. (Leave blank)
(2. SPONSORING ORGANIZ ATION N AME AND MAILING ACORESS (/nclude lip Codel p
- 11. CONTRACT NO.
13 TYPE OF REPORT PE RIOD COVE RE D (/nclusive dates)
Technical Report
- 15. SUPPLEMENTARY NOTES 14-(led"' d'"*/
- 16. ABSTR ACT (200 words or less!
This report updates and supercedes the NRC technical positions established in NUREG-0313 " Technical Report on Material Selection and Processing Guidelines l for BWR Coolant Pressure Boundary Piping," published in July 1977.
This report sets forth the NRC staff's revised acceptable methods to reduce
' the intergranular stress :orrosion cracking susceptibility of BWR ASME Code Classl,2_,&3 pressure boundary piping and safe ends.
For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented.
1; KE Y WORDS AND DOCUMENT AN ALYSIS 17a DESCRIPTOHS 1 71) 10E N TI F IE RSJOPE N E N DE D T E R YS 18 AVAILAB;LITY ST ATEMENT
- 19. SE CURITY CLASS (Th,s reportl 21 NO OF PAGES t!nt1accifind l'11imited Availability 20 SECURITY CtASS (rn,, paw 22 eRiCE S
NEC FORM 335 17 77)