ML20052D820
| ML20052D820 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/04/1982 |
| From: | Richard Anderson, Bridenbaugh D SUFFOLK COUNTY, NY |
| To: | |
| References | |
| ISSUANCES-OL, NUDOCS 8205070239 | |
| Download: ML20052D820 (60) | |
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O 8 7 1 UNITED STATES OF AMERICA
, NUCLEAR REGULATORY COMMISSION
'd? l'lh ->5 r;q .q)
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
LONG ISLAND LIGHTING COMPANY )
) Docket No. 50-322 (0.L.)
(Shoreham Nuclear Power Station, ) -
Unit 1) )
)
PREPARED DIRECT TESTIMONY OF ROBERT N. ANDERSON AND DALE G. BRIDENBAUGH ON BEHALF OF SUFFOLK COUNTY
/ N REGARDING S SUFFOLK COUNTY CONTENTION 24 (1 T gG d 3 AND SOC CONTENTION 19(c) 5 19(d) s6 a
a CRACKING OF MATERIALS AND -
MATERIAL SELECTION AT SHOREHAM ,
MAY 4, 1982 i
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SUMMARY
OUTLINE OF SUFFOLK COUNTY CONTENTION 24 '
AND SOC CONTENTION 19(c) AND 19(d)
Suffolk County contends that LILCO has not taken adequate care in the selection and control of materials used in the construction of systems and components making up safety-related systems exposed to the reactor coolant environment.
Failure of such materials through the mechanism of intergranular stress corrosion cracking (IGSCC) and through thermal cycle fatigue has been observed in nearly every BWR which has been
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in service for more than three years. This problem was first observed at BWRs in 1965 and has been the subj ect of extensive investigation. Pipe cracks have been identified by the NRC as a generic safety issue since that term was first used by the'NRC.
Two " Unresolved Safety Issues" were finally identified in the NRC's Annual Report to Congress for 1978. , These two USI's sere',
A-10, BWR Feedwater No::le Cracking, and A-42, Pipe Cracks, in ,,
Boiling Water Reactors. , ,
The NRC subsequently identified Task Action Plans on these' subjects. The Task Action Plans were concluded in 1980 with the -
issuance of NUREG-0313, Rev.1, and of NUREG-0619. With the issuance of these reports, the NRC changed the identification -
of these issues to " technically resolved" but this classification has no effect on the ultimate implementation of the technical.
recommendations. -
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3 1 s LILCO has claimed that these issues are not safety problems but rather represent potential reliability problems for future operation of the plant. Witnesses agree that the potential accidents resulting from pipe ruptures have been analyzed in the FSAR but take the position that safety is potentially affected due to the increased probability of pipe rupture, the potential problem of failure of a major reactor pressure vessel no::le which has not been analyzed, and because of the safety implications of the increased occupational radiation exposure that may result from required future modifications.
The major concern at the Shoreham plant is the use of 304 stainless steel in the construction of the reactor recirculation system. The recirculation system con ,
stitutes the majority of the piping exposed to reactor coolant and is the closest to the reactor core and therefore potentially, will attain the highest radiation levels. *The 304 stainless steel material has been observed to be susceptible to IGSCC and LILCO has only committed to perform the NUREG-0313 .
recommended inspection of the non-conforming welds in the , , ,
recirculation system to the " extent practicable". .
With regard to the feedwater no::le cracking problem, ,
LILCO appears to have committed to make the modificitions recommended in NUREG-0619. There is, however, a question as l .
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1 1 to when such modifications will be made, the completion of the recommended preservice inspection, and the effectiveness of the low flow feedwater controller that will be utilized.
Witnesses believe that LILCO could and should take further steps to assure adequate performance of these critical materials at Shoreham. Further consideration should.
be given to improving the as-built condition of the 304 piping and additional leak detection devices should be considered for . . ,
non-conforming welds. A contingency plan should be developed for the guidance of future action. LILCO's failure to take such steps shows that LILCO has not demonstrated adequate resolution of these important issues in accordance with the 1 .
appropriate regulations. ,
Exhibits:
- 1. U.S. NRC 1979 Annual Report. ,
- 2. LILCO May 15, 1981 Letter SNRC-566, J. P. Novarro '
to Harold R. Denton.
- 3. FSAR, Section 5.2.3, General Material Cobsiderations,
- 4. NUREG-0531, Investigation and Evaluat. ion of Stress. ,
Corrosion Cracking, Section 6.0.
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s PREPARED DIRECT TESTIMONY OF ROBERT N. ANDERSON AND DALE G."TEDENBAUGH REGARDING SUFFOLK COUNTY CONTENTION 24 AND 50C CONIENTIONS 19(c) 4 (c)
CRACKING OF MATERIALS AND MATERIAL SELECTION AT SHOREHAM I. INTRODUCTION
- 1. This testimony was jointly prepared and edited by Dr.RobertN.AndersonandDaleG.Bridenbrugh.h/ A statement of the qualifications of Messrs. Anderson and Bridenbaugh has been separately provided to this Board. With regard to this particular contention, the experience of Mr. Bridenbaugh is particularly relevant. He was involved for five years in the construction, startup, and maintenance of the Dresden 1 plant (Commonwealth Edison). Shortly thereafter he took over as .
manager of General Electric service for operating reactors and was responsible for the investigation and repair efforts for the pipe cracks experienced at Dresden 1 in the late 1960 's.
In this assignment he was also responsible for directing G.E. 's . . .
efforts in the correction of IGSCC failures at the Nine' Mil'e .
Point plant, the Garigliano plant in Italy, and at Tarapur in .
India. When the increased incidence of pipe cracks 'at operating reactors was observed in 1974, he was responsible for coordinating ,
1/ The inivial outline and preliminary draft of this testimony was prepared by D.G. Bridenbaugh, with the exception of Section III.D, which was drafted by Dr. Anderson. Subse-quent editing and modification was a mutual effort.
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the General Electric Nuclear Energy Division response to the NRC's shutdown order in 1974. Since leaving General Electric, he has continued to follow this generic problem and has sub-mitted testimony in the Black Fox construction permit case 1
and has worked on two studies of this problem for the Swedish Nuclear Inspectorate. Dr. Anderson is eminently qualified in the field of materials, has served as a consultant to the Electric Power Research Institute, and to the California Energy Commission. Dr. Anderson is a Professor at San Jose State University and teaches graduate courses in Corrosion
_ . Engineering.
II. STATEMENT OF CONTENTION
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- 2. The purpose of this testimony is to address Suffolk County Contention 24 and SOC Contentions 19(c) and (d) as ad-mitted by the Board as follows: ,
SC 24: CRACKING OF MATERIALS .
Suffolk County contends that LILCO has not demonstrated, and the NRC Staff has not
- verified, that Shoreham meets the require- ,
ments of 10 CFR 50, Appendix A, GDC 4, 14, 30, and 31, with regard to the adequacy of
- material selection and control and system '
design as follows:
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(a) The use of appropriate materials and pro- l cesses as specified by NUREG-0313, i Revision 1, has not been fully followed ,
in the design and construction of the '
Shoreham piping systems important to safety.
(b) Recommendations contained in NUREG-0619 (p. C-12) relating to the installation of a low flow controller to be used to con-trol feedwater flow over a range of flow from 0.5 percent to 10 percent of rated flow has not been implemented at Shoreham.
Analytical evidence shows that such a flow controller is necessary to limit crack growth in SWR feedwater no::les over the life of the plant.
SOC 19(c): REG. GUIDE 1.31: FERRITE CONTENT IN STAINLESS STEEL (c) Regulatory Guide 1.31 -- The control of ferrite content in stainless steel weld metal by LILCO complies with Revision 1 of the guide rather .
than Revision 3, with regard to . verification of delta ferrite content of filler materials and to examination for ferrite content by a magnet-ic measuring instrument. Therefore, Shoreham .
does not comply with 10 CFR Part 50, Appendix A, .
Criteria 1 and 14 SOC 19(d): REG. GUIDE 1.44: MATERIALS - -
(d) Regulatory Guide 1.44 --
LILCO has not ade-quately demonstrated control of the applicati.on . ..
and processing of stainless steel to avoid severe sensitization that could lead to stress corrosion cracking as required by 10 CFR Part 50, .
Appendix A, Criteria 1 and 4 and 10 CFR Part 50, ,
Appendix B in that Shoreham has failed to. comply with the NRC Staff position described 'in NUREG-0313, Revision 1 as follows :*
It should be noted that LILCO's use of alphanumeric designa-tions is not in compliance with the nomenclature set forth in NUREG-0313, Revision 1. This discrepancy makes LILCO's commitment inscrutable.
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- 1. Portions of the reactor recirculation system i
(331) and stainless steel to carbon steel transi-tion welds between the reactor recirculation sys- l tem and the reactor water clean-up, core spray, and residual heat removal systems do not meet the guidelines set forth in NUREG-0313, Revision 1, for ASME code class I and II reactor coolant pres-sure boundary piping.
- 2. The commitment to inspect portions of the re actor recirculation system and transition welds
.. that have been classified as "non-conforming" per
. NUREG-0313, Revision 1, has been conditioned by LILCO to be limited "to the extent practicable" due to physical interference in some locations.
NUREG-0313 does not specifically allow for such deviations. Also, LILCO has failed _ to identify specifically the number, location, and detailed justification for these deviations. Further, LILCO's obj ection to the " service sensitive" classification of recirculation riser lines and inlet lines at the safe-end curves demonstrates a failure to comply with the requirements of i- NUREG-0313. .
- 3. The limiting conditions for leakage included' as part of the technical specifications has not been demonstrated in that the leak detection sys-tems may not adequately enhance the discovery of unidentified leakage as required by NUREG-0313. ~ ,
The results of our review of some of the important matters en-compassed by these contentions are summarized in the 'following ' .
paragraphs. , . ..
III. DISCUSSION OF ISSUES .
III.A.: Background and Summary of Position
- 3. The basic concerns expressed by these contentions deal with the potential inadequacies of the primary system ma'-
terials, particularly their selection and control, to assure retention of the reactor coolant pressure boundary (RCPB) and 4
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f 3 of the safety-related piping systems' integrity for safe plant operation. Cracking of stainless steel materials used in critical piping systems has been a generic problem with boil-ing water reactors since the mid-1960's. Cracking of a by-pass line in the Dresden-1 reactor recirculation system was observed in 1965 and investigation of this problem and work -
, to develop suitable alternate materials has been an industry wide concern for the past 15 years. BWR piping system crack-ing was identified as a generic safety issue on the NRC's first generic issue list issued in 1977, and in 1978 this issue was reported by the NRC to Congress as one of the 17 unresolved safety issues affecting existing and future boiling water re-actors. With respect to the Shoreham Operating License appli-cation and this contention, two unresolved safety issues were '
identified by the NRC which are of relevance. These are:
A-10, BWR Feedwater No::le Cracking,and A-42, Pipe Cracks in Boiling Water Reactors ,,
Copies of the summary sections of the 1978 NRC Report to Con- .
gress identifying these issues as unresolved safety issues are appended as Exhibit 1. *
- 4. Following the NRC's identification of the A-10 and ,
A-42 unresolved safety issues, these prob'_ ems were given top -
priority within the NRC and technical solutions and/o'r improve-ments were developed and documented in two NRC reports. !
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2/ NUREG-0313, Rev. 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping", July,1980 and NUREG-0619, "Feedwater No::le and Control Rod Return Line Cracking", November, 1980.
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f g These reports, issued in 1980, formed the basis for the NRC to declare these problems as " technically resolved", and this position has been so documented in the NRC's report series on unresolved safety issues.2/ A number of points must be em-phasized with regard to this " technical resolution":
(1) The technical advances were not made in the -
confines of the NRC's Task Action Plans.
The NRC's Task was limited to the review and approval of material development that result now from years of nuclear industry research and development. NUREG-0313, Rev. 1 formally recommends implementation of improvements that have been under in-vestigation for a number of years.
(2) " Technical resolution" means only that the NRC believes the technology exists to re-solve the problem and makes no j ucgment as to whether it has actually been imple-mented at existing plants.
With respect to implementation, NRC plans were that c,ommitment' letters would be solicited from each licensee with respect to the two issues and that these letters would be reviewed by the Staff and decisions would be made as to their adequacy. .A ,,
letter with respect to these two issues has been submitted by .
LILCO to the Staff (See LILCO letter SNRC-566, dated May 15, 1981. A copy of this letter is appended as Exhibit 2.) As -
will be discussed in later sections of this testimony, the com- ,
mitments made by LILCO do not assure complete adherence of the
- Shoreham Nuclear Plant to the recommendations containe'd in the NRC's implementation documents.
3/ NUREG-0606, Vol. 3, No. 2, Unresolved Safety Issues Summary, May 15, 1981.
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- 5. An additional discrepancy in the LILCO material program is failure to include improved methods in the con-struction of the plant. One such improvement is use of advice contained in the latest revision of Regulatory Guide 1. 31, Ferrite _ Content in Stainless Steel. Our opinion is that the failure to fully implement the improvements identified in '
these three documents (NUREG-0313, NUREG-0619 and Regulatory Guide 1.31, Rev. 3) represent deficiencies that result in the plant not being in compliance with the appropriate regulations; namely, 10 CFR Part 50, Appendix A, Criteria 1, 4, 14, 30, 31 and that they further serve to increase the risk to the public health and safety.
III.B: Industry Experience
- 6. As indicated in previous sections of this testimony, the problem of pipe cracks in safety-related piping systems in boiling water reactors has been a serious problem nearly since the beginning of BWR commercial reactor operations. Even prior .
to the identification of this issue as an unresolved safety
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issue by the NRC in 1978, numerous failures experienced at BWR's, both inside and outside the United States, had caused .
the NRC to activate a task force as early as 1974 to study the' potential safety impact of these failures. This task force, known as the Pipe Crack Study Group (PCSG), was formed by the
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l NRC following a number of failures experienced in 1973 and 1974. I The PCSG published its findings in a 1975 report, NUREG-75/067,
" Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants", published October, 1975. In additon to the PCSG effort, all BWR's in the U.S. were required to be shut down in late 1974/early 1975 for mandatery.
piping inspections to assure that undiscovered cracks did not .
exist. The b'asic finding of the first PCSG review was that inter-granular stress corrosion cracking (IGSCC) is a relatively com-plex, progressive cracking mechanism that occurs under the si-multaneous presence of three conditons: high stress; sensitized material; and a corrosive media. In the BWR application, welding of the commonly selected piping materials provides the stress and the sensitization, and the relatively high oxygen levels'in-herent in the BWR coolant provide the corrodent. This early PCSG report also found that this issue, although a serious re-liability problem, was not considered to be a safety problem b.e-cause experience to that time had favored the so-called " leak -
before break" theory.
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- 7. In June, 1978, a through-wall crack wa's discovered 'in an inconel recirculation riser safe end at the Duane Arnold ,
P1 p.n t . Discovery of this leaking crack was the result of a rou'-
tine walk through inspection of the drywell following a shutdown for other reasons. The leak detection system had not given I
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indication of a leak during operation. Through wall cracks were found in the 10 inch riser and extended around 270 degrees of the piping circumference. Furthermore, it was found that cracking was present in the other seven recirculation risers in the same location. This event gave rise to concern over the random event theory ascribed to such cracks and also brought
, question to the ability of the leak detection systems in use to detect cracks before potential full rupture of the pipe itself.
A second event observed at a German BWR designed by General Electric was the discovery of sensitized safe end cracking in large diameter (24") pipes. The combination of these two events caused the NRC to reactivate the PCSG in September, 1978. The PCSG studied the issue for approximately one year and its find- '. ..
ings were published in a 1979 report, NUREG-0531, " Investigation and Evaluation of Stress Corrosior Cracking in Piping of Light Water Reactor Plants", published in Februgry,1979. The data assembled during the PCSG investigation formed the technical ..
basis of NUREG-0313, Rev. 1. -
- 8. In addition to the piping failures experienced at operating BWRs reported in NUREG-75-067 and NUREG-0531, .
other material failures have been experienced at U.S. BWRs .
These include cracking of vessel feedwater no::les,, cracking of internal feedwater spargers, cracking of control' rod drive return no::les, and cracking of a core spray sparger. Details
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of the vessel feedwater no::le and control rod drive return line no::le cracking are contained in an NRC report, NUREG-0312, " Interim Technical Report on BWR Feedwater and Control Rod Drive Return Line No::le Cracking", published in July, 1977 The feedwater no::le cracking that has been experienced at almost all U.S. BWRs which have been in oueration several'
. years is of the most significant conce rn . Safety analyses supporting the licensing of BWRs do not consider the possi-bility of vessel rupture. Cracking of the feedwater no::les, an integral part of the vessel, raises questions on the wisdom of such assumption. The possibility of such a crack growing into a vessel rupture case, the extreme difficulty of repair-ing such defects, the high radiation exposure associated with .
the repairs, and,the possibility that irreparable cracks cou'Id be generated which would result in the need for early retire-ment of the unit serve to emphasize the sa.fety significance of.
these material failure issues. Correction of this problem ..
(feedwater no::le cracking) was the subject of intensive work -
by the industry and by the ' BWRs designer, General Electric.
The NRC report, NUREG-0619, presents the latest state o'f tech-nical knowledge in this problem area and recommends usage of .
the latest design thermal sleeve (a triple sleeve) plus other operating and inspection procedures. LILCO has committed to utili:e the triple sleeve feedwater sparger design at Shoreham, l
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but LILCO's documentation does not state if this modification will be completed before initial operation. LILCO also com-mits to the installation of a low flow controller, but waf-fles on whether this controller will actually meet the re-quirements specified in NUREG-0619.
- 9. Regulatory Guide 1.31, Ferrite Content in Stainless Steelj was issued to assure proper control of ferrite in stain-less steel weld metal as installed. The latest version of this Regulatory Guide, Revision 3, requires verification of delta ferrite content of filler materials to be measured by a mag-netic measuring instrument. This would provide additonal as-surance that the as-welded conditon is as desired and of less susceptibility to cracking. LIL.C0, however, has committed in the FSAR only to comply with Revision 1 of the Guide which does not require this additonal step. In view of the demon-strated susceptibility of boiling water, reactor piping spstems to numerous cracking failures, we believe it is essential 3 hat the latest technical advances be utili ed and that this addi-tional test should be performed.
III.C: Shoreham Commitments ,
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- 10. LILCO commitments with regard to assuring compliance with the latest technical recommendations for the RCPB and safety related picing systems in the use of crack resistant materials l
- , e is contained in Section 5.2.3 of the FSAR and is supplemented by LILCO commitment SNRC-566. The position stated in the FSAR is as follows:
"The use of severely sensitized or furnace sensi-tized stainless steel is prohibited. However, exception is taken to Regulatory Guide 1.44 which effectively prohibits the use of welded pipe 304 stainless steel by considering it the same as '
severely sensitized material and by placing limits on oxygen concentration which cannot be met by operating BWR's." (A copy of this FSAR section is appended as Exhibit 3) .
LILCO supplements this position in the May 15, 1981, letter SNRC-566 and states that "the problem of stress corrosion crack-ing has, to a large degree, been eliminated for Shoreham".
LILCO points out in this letter that the recirculation bypass line has been eliminated, that .the core spray line and safe end materials have been changed and the CRD return line has been eliminated. This then leaves the reactor recirculation system as the major primary pressure boundary piping which utilized 304 stainless steel material.d/ LILCO states that modifica- ..
tion of the piping material for this system to an alternative -
material was not practical in the time table for Shoreham. Un-fortunately, the 304 reactor recirculation system represents by far the majority of the piping (in terms of internal surface .
exposed to reactor coolant) and, therefore, still provides a '
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~ LILCO's submittals in repsonse to NUREG-0313, Rev. 1 are silent on RHR/LPCI, and auxiliary piping system connection ; materials.
These systems admittedly are a small percentage of the RCPB when compared to the recirculation system, but their exact material status is not identified. )
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major opportunity for the initiation of IGSCC failures. We have not attempted to exactly quantify the piping system per-centage of the wetted area provided by the recirculation system.
It should be emphasized, however, that the recirculation sys-tem is the largest and most complex piping system making up the nuclear system. NUREG-0531 contains a good summary description of the piping systems making up a BWR. A copy of Section 6.0 of this report is appended as Exhibit 4 Note in particular Figure 6.1 which shows the complex discharge header and riser i configuration. Section 6.2.2 compares the number of welds con-tained in a typical BWR recirculation system to the number in the core spray piping. The recirculation system contains 132, while the core spray system has 54 However, if the pipe dia-met,ers are taken into account, we find that the recirculation system contains about three times the lineal footage of pipe welds that the core spray system does. LILCO attempts to justify the 304 piping in the recirculation system by claims ..
that field weld improvement measures were utilized, including -
welding process control, grinding restrictions, and weld filler material ferrite control as required by Requlatory Guide 1.33.5/ '
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5/ Regulatory Guii' 1.33 is Quality Assurance Program Requirements, Operation, so LILCO in all probability was referring to Regu-latory Guide 1.31, Control of Ferrite Content in Stainless Steel Weld Metal. LILCO's apparent uncertainty in exactly which Regulatory Guide they followed and their failure to identify the revision number (there are three revisions of 1.31) does little to assure confidence in the program as
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LILCO further commits in SNRC-566 (Exhibit 2) that recircula-tion system welds and transition welds classified as "non-conforming" will be inspected in accordance with Part III.3.1.0 of NUREG-0313 to the extent practicable for the Shoreham Nuclear
, Power Station (emphasis added) .
- 11. LILCO's commitment with regard to feedwater no:rle
, hardware is also contained in SNRC-566. This letter states that the General Electric triple sleeve sparger has replaced the welded-in design. It also states that performance of a baseline UT will be performed after installation o# the sparger. However, the Safety Evaluation Report (NUREG-0420, Supplement 1) states the Applicant has committed to install the triple sleeve sparger.
SNRC-566 further states that the baseline inspection will be '
performed after the installation of the sparger. No date is stated for sparger installation. While there is no doubt that LILCO intends to install the new sparger,,it is not clear that it has yet been done, nor that it will be done before plant oper-ation. LILCO also states that a low flow feedwater controller -
(0.1% capability) has been installed, but that monitoring at startup could determine a need for " additional controls'." This implies the use of procedural controls. This does not meet the .
guidelines of NUREG-0619, which call for the use of an accurate l
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III.D: Material Selection Uncertainties
- 12. Stainless steel type 304 piping was specified in early BWRs as the material to be used in most of the safety-related piping systems exposed to reactor coolant. This mater-ial was selected for this application until very recently, par-
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tially because of its common availability and also because it
. was thought to be an acceptable material for this application.
Cracking (IGSCC) experienced over the past eight years had led to the conclusion that 304 is unsatisfactory for use in BWR coolant systems unless it can be left in the so-called unsensi-tired condition so as to guard against cracking and failure.
Some improvements have been made in installatica techniques, but the field welding process makes total control nearly impossible.
The following paragraphs in this section (III.D) fur ther exp1'ain, the nature of the 304 failures and demonstrate that the uncer-tainties in prevention and early detection .of failures makes re-placement of this piping with a more suitable material de. Abke.
- 13. Experience and studies to date have shown that the.
combination of three conditions , high stress , sensitized mater-ial and a corrosive media, will produce IGSCC. IGSCC of austen- ;
itic type 304 stainless steel originates in the heat affected zone , l (HAZ) adj acent to a weld. The welding operation causes heating and cooling of the metal through the range of 870-425 C which, in turn, produces chromium carbide precipitate at grain boundaries. I This precipitation reduces the chromium content available fbr a .
resisting chemical attack at grain boundaries. These altered grain boundaries are referred to as sensiti:ed because corro-sion can penetrate deeply and quickly along the chromium de-pleted boundaries if high stresses and an aggressive environ-ment are present. Since all three of the above mentioned con-ditons must be present to some degree for IGSCC to occur, ir
, is theoretically possible to avoid its initiation by eliminat-ing one or more of the three conditions.
- 14. Avoidance of sensitization can diminish or minimize the onset of IGSCC by using low carbon (less th'an 0.04 weight percent carbon) material. Another successful method has been the use of corrosion resistant cladding on the inner surfaces of the 304 pipe material so as to isolate the sensitized re-gions from the aggressive environment.
- 15. The elimination of high stresses or optimal control of stresses is the second possible solutiqn to the problem. Re-lieving of residual stresses caused by the non-equilibrium cool-ing after welding can be done by the use of heat sinks on the in-terior of the pipe during welding or by induction heating after welding. There is, however, no practical means at the'present time to verify piping stress reduction to a sufficiently low .
level to conclusively prevent IGSCC from initiating.in 304 material.
- 16. The third potential area for solution of this prob-l lem is to control the environment under which the material l
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operates so as to remove the corrosive mcdia. Dissolved oxygen in the BWR coolant at a level of 0.2 ppm can act as a corrosion catalyst if sensitization and tensile stresses are present. The presence of oxygen in BWR water is an inherent characteristic of ;
the process and control of it to the degree necessary to protect ;
the sensiti:ed pipes does not appear to be feasible at this time.
- 17. Certain procedural controls can serve to retard the onset of IGSCC. Limiting grinding on the inner pipe surfaces and prevention of microfissures by control of ferrite in the weld helps to prevent crack nucleation. Unfortunately, the efficacy of such controls cannot be readily verified, and the safest ap-proach remains to use nuclear grade materials that are not readily suscentible to IGSCC.
- 18. Numerous studies have been performed and continue to be under way in an attempt to identify the best alternate material to use so as to minimi:e or prevent IGSCC.. Studies have identi-fied a number of materials that are more suitable in 55R applica-tions where conditions conducive to the promotiom of IGSCC are -
present. One material that has been identified and used is 304L (low carbon) stainless steel. This material is commonlp used in industry for such service and has been used as replacement mater- .
ial for many of the failures experienced at BWRs to .date. 304L' has 85% of the yield strength of 304 and can be substituted di-rectly for 304 in most applications. Its primary drawback is increased cost. In the past it has generally been less avail-able in the commercici market but that situation appears now to be changing quite rapidly.
- 19. Certain of the studies conducted in recent de-velopment programs have focused on the development of tech-niques to measure 304 sensitization so as to permit identifi*
cation of as-installed piping systems that will be more sus-ceptible to IGSCC. The electrochemical potentiokinetic re-activation (EPR) method is one that appears promising. The EPR method measures the current flow during a potential sweep and compares this measurement to a similar measurement of flow in non-sensitized material. The results of this measurement correlate to the actual rate of corrosion in the grain bound-ary when corrected for grain si:e. The EPR method appears to' effectively identify sensiti:ed materials and is useful for quality control but it has not been proven , effective in the field construction application where electrolyte, stress level,,
and grain size are difficult to control. It is possible that .
EPR measurement could give some indication of the liklihood of IGSCC 'at Shoreham, but a conclusive outcome from such a ' test is unlikely. ,
- 20. Testing for the presence of IGSCC in existing pip-ing is difficult because of the nature of the cracks. The cracks are small and tight with many branches and ultrasonic testing 1
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and radiography techniques have some problems in the unequivo-cal identification of cracks. Moreover, once detected, crack si:e and hence, severity, may not be readily identified. Acous-tic emission using the sound of deformation or cracking in mett.1 is one possible detection technique, but this method is hampered by the problem of background noise. '
- 21. Two arguments that have been advanced to support the use of sensitized 304 stainless piping in reactor applica-tion such as the recirculation system are (1) the leak-before-break argument which assumes that IGSCC will propagate asymmetrically and exit the pipe to produce a leak before pro-ducing a significant loss of strength in the pipe. The leak then would be detected before the pipe crack propagated far enough to leak in the pipe where it fails under normal service .
loads and (2) smaller pipes are more likely to fail by IGSCC than larger pipes. The first argument ignores the very real possibility that multiple cracks may occur in the same RAZ. ..
which would rapidly weaken the pipe before any signs of leaking -
might occur. Moreover, some experts have pointed out as quoted in NUREG-0313, Rev. 1 that: -
"The residual stress distribution in circumferential .
welds tends to promote cracking all around the cir-i cumference. As the crack extends around a signifi-cant portion of the inside wall, the residual t
stresses in the axial direction should increa'se and accelerate the crack growth." 6/
6/ NUREG-0313, Rev. 1, " Comments" by Lawrence Livermore Lab. ,
l Finally, investigators have stated that differences in the sus-ceptibility to IGSCC between large diameter and small diameter 304 pipe may be due only to heat-to-heat variations in the ma-terial. It is clear that crack propagation rate and the as-sumption that crack growth is asymmetrical with the walls of the pipe are only hypothesis at this point and the leak-before-break argument can be considered a shibboleth.
IV. CONCLUSIONS
- 22. LILCO's primary reason for retention of the 304 ma-terial in the Shoreham recirculation piping appears to be that the cost of such modifications would be excessive and that chang-ing of the system to an alternative material "was not practical in the time table for Shoreham".2/ In our opinion, the con-tinued use of type 304 stainless steel with its limited life, difficulty of detecting failure mode, and the possibility that repairs will exacerbate the cracking means a conmitment to,futyre detection and repair that can have adverse consequence.s for the ,
safety and economics of the Shoreham station. It is for these reasons that susceptible piping.should be replaced with material that will be less adversely affected by the oxygenated water. ,
We recognize, however, that these piping systems are already con- ;
structed, that the cost and schedule delay associated with full 2/ LILCo letter SNaC-s66 (see Exhibit 2). ,
replacement would be substantial and that it may be impractical or impossible to effect such a change at this time. We believe that those sections of recirculation system piping that may be considered as "high-risk" weldments that cannot be replaced should be equipped with state-of-the-art (solid state) leak detector elements in the welded areas and backed up by appro-
. priate acoustical detectors when available. The "high-risk" weldments should be identified through the use of a careful study of piping and electrode material composition, review of stress levels at each joint, and augmented by utilization of
~
EPR and delta ferrite measurements.
- 23. LILCO further should identify all welds in these systems which are not fully available for ASME Section XI in-service inspection. Special attention should be given to leak detection at these joints. We also recommend an evalua-tion of accidents resulting from possible . failures at these locations to supplement hazard analysis already contained in ..
the FSAR. *
- 24. Based on our review and discussion summarized in SEction III above, we conclude that: -
(1) LILCO has not demonstrated full com- .
pliance with NUREG-0313, Rev. 1 and, accordingly, does not meet 10 CFR 50, Appendix A, Criteria 1, 4, 14, 30 and l 31. Deviations include extensive use l
- of 304 stainless piping materials and r
l
failure to apply augmented inspection to all service sensitive applications.
(2) LILCO has not complied with Regulatory Guide 1.31, Rev. 3. Compliance with this Regulatory Guide could add assurance of quality construction of the critical -
, piping systems.
(3) LILCO has not committed that the triple-sleeve sparger as called for in NUREG-0619 will be installed on a timely ,
basis. They have further not committed that an effective low feedwater control-1er will be provided at Shoreham.
- 25. We recommend that:
(1) The requirements contained in NUREG-0313, Rev.1 should be fully followed at Shoreham.
This means that the 304 material in the . ..
in the recirculation system should be .
replaced with an alternate material or all welds in the system should receive appro-priate post-weld treatment or should be .
protected from the coolant environment.
(2) If recommendation (1) above is not' fully completed with, LILCO should be required j' to fully justify all deviations from the NUREG document. Such justification should include:
(a) Issuance of a full report, including isometri.c piping drawings showing the ,
locations of all 304 piping systems exposed to reactor coolant. All weld locations should be indicated as should all areas where access or other conditions prohibit full compliance with ASME Section XI inspection requirements.
(b) A feasibility study should be performed to determine if any of the above identi-fied piping can be made more resistant to the onset of IGSCC through the use of CRC, IHSI, or other post-installation methods.
(c) A contingency plan should be prepared addressing the cost, schedule, and availability of alternate piping ma-terial in the event of post-operational failure of the 304 stainless piping systems. The plan should also quantify occupational radiation exposure required for the complete replace-ment in the future, at various times in plant life.
(d) LILCO perform measurements of deposited weld metal to bring the as-built systems into compliance with Reg. Guide 1.31, Rev.
3, to the extent possible.
(e) Apply the EPR test to the installed system to the extent possible to obtain the -
possible benefit of this test.
(f) Investigate the use of advanced leak detection methods such as the high- -
sensitivity leak tapes (these are now being developed under an NRC contract) to ensure adequate detection at all non-conforming j oints and locations .
(g) Document'and make available a summary report showing that all 304 stainless piping material certificates verify the level of carbon and cobalt material in the RCPB piping systems . Utilization of low l l
1evels will demonstrate lower probability l
l !
l
of failure and reduce the likelihood of higher radiation exposure rates.
We additionally recommend:
(3) LILCO should demonstrate that no furnace-sensiti:ed material remains in the RCPB (via ,
assembly of records , tests , etc. as needed) .
~
(4) LILCO should provide verification that all reactor internal components constructed of 304 stainless material are in the solution heat treated condition and conform with the provisions stated in FSAR Section 4.5e e
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I .hti The President I ) % *
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- "' The White House I Wmhington, D.C. 20500 x. .M-vg;F i
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i: 997..?ses- .
Dear Mr. President:
Enclosed is the fifth Annual Report of the United States Nuclear Regulatory Commission for your transmittal to the' .
Congress, as required by Section 307(c)'of the Energy Reorga-nization Act of 1974. This report covers the major activities of ;
the NRC from October 1,1978 through September'30,1979'and' briefly desc4bes some additional actions through 1979 into 1980. .
e Respectfully, I
Y
~
w b John F. Ahearne
i . ,
- 73 I, II, and III contairw) nt designs. The results b Three MileofIsland event. scenario indicated that a Centric Task A-39 willy an integral part of the final number of aspects of thkATWS accident evaluation acceptability of these tsigns. The portions of this required reconsideration h! specially for PWRs. l 3 gtn ric task related to .h'e Mark I and Mark II con- The shortage of avIilable industry manpower l tainments are currently :cheduled to be completed in delayed several of the reqdred submittals of confirma- i March 1980. tion analyses. The resultdas a substantial slip in the I projected completion da(cifor the ATWS task. NRC staff manpower was partipy restored in June 1979 Anticipated Transier laI t Without Scram and meetings were held wah industry representatives qe in July and August of 197 Fo discuss the impacts of the Three Mile Island Unit Qaccident on their respective
. Nuclear plants have sqfoty and control systems to ATWS evaluations. As cfSanuary 1,1980, the NRC
. limit the consequences o bnormal operating condi- staff planned to propose d ATWS rule to the Commis-tions transients. Some p* H deviations from normal sion by April 15, 1980, i :h a goal of issuing a final operating conditions ma less frequently, mayose im@pe minor; others, ATWS rule by Decem occurnng 980. .
significant demands on plent equipment. " Anti ciiatedt Operational Occur-rences" or " Anticipated dansients" are defined (10 BWR Nozzle Cracking; ,
CFR Part 50, Appendix Afas "those conditions of nor-mal operation which ar xpected to occur one or Over the last several years, inspections at 22 of the mora times during the lif gf the nuclear power unit 31 boiling water reactor (BWR) plants licensed for and include but are not ! n operation in the U.S. have disclosed some degree of
' gted to loss of power to all recirculation pumps, tnp upg of the turbine generator cracking in the feedwater nozzles of the reactor vessel set, isolation of the maingondenser, and loss of all off- at 18 facilities. One facility has not yet accumulated significant operating time and has, therefore, not yet
, sita power. In some antigipated transients, rapidly- g g shuttmg down the nuclear reaction (initiating a scram ), and thus rapid y ucing the generation of The feedwater nozzles are an integral part of the heat in the reactor co ;e, is an important safety primary pressure boundary of the reactor coolant measure. If there were a potentially severe anticipated system and the second barrier (after the fuel cladding) to the release of radioactive fission products. All of the transient and the reactoQhutdown system did not repaired BWR feedwater nozzles met the ASME.
function as designed, then an anticipated transient.
pressure vessel code limits, however, and no im ,
without-scram, or AT%S.;would have occurred. mediate action was necessary. Because only relatively This issue has been diqcyssed throughout the NRC small amounts of metal have been removed by repair and AEC and the nuclear l industry for a number of operations, there has been no significant reduction in ysars. Details on the safety significance of the issue safety margins. Nevertheless, the erack6g is potential '
and actions taken by NRC and industry prior to fiscal ly serious because:
ytar 1979 in response to Ltimay be found in the 1978 -
NRC Annual Report, pp37 and 28.
- Excessive crack growth could lead to impairment On the basis of disdssions with senior NRC of pmssure vessel safety margins. .
management, the Adviss Committee on Reactor
- The. design safety margin could also be reduced-representatives, and the by excessive removal of nozzle reinforcement Safsguards, revisw of the Lewis and industjylittee report on the Reactor Corrrq while grinding out cracks, and repair by welding Sefity Study, the NRC stiflin December 1978 propos- would be complicated. -
ed a combination of pr9ventative and mitigative
- The exposure to radiation of the personnel ,
means of providing protect, ion from ATWS events. In P"'A"'ing inspection and repair tasks can be this supplement, the N3$ staff proposed different considerable. ,
types of plant modifications. The design alternatives
- The repair of these kinds of cracks can result in which were proposed takd ihto consideration the status considerable shutdown time at the plant affected. ;
of the plants-whether o n! rating, under construction The reactor vendor (the General Electric Company) l or nearly ready for opera :lon-and questions of prac- and the NRC have concluded from their separate tietbility, including the oost of such modifications, studies that the cracks are initiated by rapid 'fluctua.
l In order to confirm tha[ these alternatives provided tions in water temperature on the inside surface of the tha needed level of safety, tHeindustry was required to nozzles during periods of low feedwater temperature provide the necessary corfi'rmation analyses and the when flow may also be unsteady and perhaps inter-staff originallyintended tc hakeits recommendations mittent. The cracks then grow deeper as a result of to the Commission in the fring of 1979. The Thre.s operational startup and shutdown cycles or other Mila Island Unit 2 accid i affected these plans in operationally induced transients. The stainless steel
! several ways. First, both in very and NRC staff man- cladding exhibited less resistance to crack initiation power were diverted from TWS work; second, the than the underlying low-allow steel.
74 ;-
k FEEDWATER NOZZLE i
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I Cr:cks in noules of the facdwater and control rod driva lines of indicates that abrupt and wide fluctuations in water temperaturcs uwn reactor prenure veswl have been studied by the vunder (see die m above) are the initial causes of cracking. Photo above (General Elvetrict and tfw NHC staff for several years. Evidenou shows : cracks.
, l t
- I The vendor has performed extensive analysis and The staff has now completed its review'of the 8 testing to confirm the suspected cause of the cracking General Electric studies on feedwater nczzle cracking . ,
and to develop possible long term solutions-a newly and has concluded that the new sparger design-in ,
designed sleeve, removal of the stainless steel ciadding, conjunction with other remedial measures, such as i reduction of the temperature differential at the nozzle, elad removal and more appropriate operating pro-or some combination of these. The licensees involved cedures-is an effective means of greatly reducing the l'
htve increased the number and extent of in-service in- probability of crack initiation. The new sparger design i rpections of feedwater nozzles, with careful repair and includes flow discharge nozzles, a triple thermal ;
reinspection where cracks were found. The vendor ad- sleeve, and two piston ring seals in the nozzle bore. I vised these licensees to closely monitor startup and The effectivene:s of the new design in minimizing by- ,
shutdown procedures in an effort to substantially pass leakage and other problems encountered in the-- ;
reduce the time during which cold feedwater is being older designs was confirmed by extensive testing and ,
injected into the hot pressure vessel. analyses by General . Electric, including vibration, '
In a closely related area, the NRC was informed in thermal-hydraulic, materials,' and thermal fatighe March 1977 by the General Electric Company that a evaluations. Other designs may alsote acceptable. ..
cr:ck had been found in the nozzle of the control rod Feedwater system changes, necessary on some low !
drive (CRD) return line in a reactor vessel. The CRD feedwater temperature plants to assure no cracking
- return line nozzles are the openings in BWR pressure over the lifetime of the plant, are being evaluated on a i vessels through which the high pressure water in excess plant specific basis. An NRC staff report incorporat- i of that needed to operate and cool the CRDs is return- ing guidance for operating reactors and plants under I cd to the pressure vessel. The cracks resembled those licensing review is in preparation and is scheduled to '
fcund in the feedwater nozzles and seemed to be the be issued for comment in February 1980.
result of the same kind of cyclic thermal stresses that The resolution of questions regarding the future w:re causing feedwater nozzle cracks. The maximum selection of improved inservice inspection techniques crack depth has been 0.87 inch. and frequency of inspection has been separated from The NRC staff efforts related to the resolution of the generic task while major industry investigations these two similar issues regarding nozzle cracking in continue -(including thermal cracking in a full-size i b:iling water reactors were consolidated into a single nozzle mockup to be used in ultrasonic evaluation). A '
staff effort, Generic Task A 10, in 1977. Under supplement to the NRC staff report cited above may '
C neric Task A 10, the staff issued interim guidance to be necessary upon completion of these studies. In the '1 operating plants in a report entitled, " Interim meantime, stringent inspection requirements, based Technical Report on BWR Feedwater and Control l mainly upon dye-penetrating testing, are stillin force. ;
Rod Drive Return I.ine Nozzle Cracking," in 1977. All licensee efforts, such as system and operational !
l
'75 chenges, to lengthen the time to crack initiation and to tion of pressure and terr rature will remain well slow crack growth are taken into account in the deter- below that which might ense brittle fracture of the mination of inspection techniques and criteria. , reactor vessel if a significa E flaw were present in the The CRD nozzle issue will be resolved by a com- vessel material. The effect d neutron radiation on the binction of actions which includes nozzle inspection fracture toughness of the vAsel material is accounted
' end repairs and some CRD system notifications. Cer- for in developing and revis dg these technical specifi-
, tain system modifications recommended by General cations over the life of the t nt.
Electric involved cutting and capping the nozzle and For the service time i operating conditions rsturn line but that action would reduce the capability typical of current operating ants, reactor vessel frac-to direct high pressure water through the CRD system ture toughness provides ac ' juate margins of safety, whtn the vessel is otherwise isolated. Although this against vessel failure. Fur r, for most plants the systsm is not normally expected to perform this fune- vessel material properties ai such that adequate frac-tion in safety analyses, the capability played a major ture toughness can be main ined over the life of the rols in keeping the core covered during the incident at plants. However, results from a reactor vessel Browns Ferry Unit 1 on March 22,1975. As a result of surveillance program indica:e that up to 20 older -
its review of these modifications, the NRC has con- operating pressurized wate* reactor pressure vessels -
cluded that only a limited number of plants will be were fabricated with m terials that will have allowed to modify the CRD system in accordance with marginal toughness after codparatively short periods ths CE recommendations. Unless the licensees of the of operation. This issue has been incorporated in the remaining plants demonstrate, by testing, that suffi- j cient flow is available to the reactor vessel with the ;
return line removed, they will be required to retain ,
the return line, rerouted to the feedwater line or a l
similar suitable connection that doesn't have the i potzntial for cracking in the reactor vessel nozzle. The ~ ir p-staff's evaluation, conclusions, and guidance on the .
c"Ik y.%d i ['
- 3. M:. NNyaSL CRD return line nozzle issue will also be included in the February 1980 NRC staff report referred to above. U ",U h Q
=
W; *h Plant-specific implementation of the generic licens- .h f p3$ !4 ,
M,,;.M.I'%, f y hj ing positions developed under this task (with the ex.
ception of future inservice inspection questions) has (d ; .
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.i-alrezdy begun.
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- ; L.., .
i.h.yy{q,9.Q:: L ." k, \
Reactor Vessel M terial Toughness '
o g .~x 1 "
g5 Nuclear reactor p re vessels are required to have en adequate margin 1 protection against fracture in b MifbM -% h* h h .h
- M
- S ' .!'M Q]. vadN_I/b d(,.MD O
tha presence of relativ } large postulated flaws. This requirement is impose f
E $e i ( dy..y l4dd
""%,s .q'd ;j evIn though extensive,jor the sake of conservatism, periodic inservice inspection b 7 "
D }
'P-programs serve to p vide protection against the presence of such fla d Fracture mechanics-the 45 ' -
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[
1 angineering method us% to establish the failure e i M. ..pg 4 r % d l margin-employs a qua 4titative material property I
[re- g - -
c:lled fracture toughnes .'to calculate the conditions e ;f ; '%g;;-[ te '.' '
und:r which catastrophicdly rapid crack propagation will occur. Fracture tougbiess has different values and 1;
4(
b.g gl lf . @ y charceteristics dependind ;upon the material being 4
qqu g 7 t '7<
.i; p
['v A .. i considered. For steels useWin nuclear reactor pressure vess:ls, three facts are iinportant. First, fracture N"4 H .'.' ;<?.' !,j,
- F toughness increases with .ncreasing temperature. Se-Tibg %
(-
. ,j [
cond, fracture toughness ' decreases with increasing > 3.kJ.,sdjyg 3g 7 ? C/r, 'T. r I- 1 '
lord rates. Third, fractu W toughness decreases with neutron irradiation.
I UA pM61
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In recognition of these qsiderations, the technical The pr ein,. in i. tion b.s b n pui O .sia. rollowing th.
specifications for power hactors set limits on the, tastes of a wend- air portion of a su thick pr==sur. ves==L operating pressure during\ heatup and cocidown A flaw mor. th n7m inches deep andf13 inches long was created
]
in t3. ., which ..s th.n ,W.cied u - d ds mor.
1 operations. Thane restrictioMassure that the combina- then doubi. th design pressun withe aisruptive tailui
1 . ,
S2 I l
i for the movement cheavy loads over spent fuel to The NRC Office of Nu'eln Regulatory Research is assure that the potential for a handling accident that could result in danMe of spent fuel is minimized also undertaking a relatedy put more comprehensive and long-term program p develop mathematical while the generic evaluation proceeds. In addition, the models to realistically pryict the probability of liernsees were req'.45{ed to provide information on radiactive releases from sei rmically induced events in loed handling operntJons for use in the Task A-36 nuclear power plants. Thid Seismic Safety Margin review. Responses wWe received from alllicensees by Research Program will utiliis input from Task A-40 in December 1978. 1 a number of areas. N The staff has comIleted its survey ofload handling Ceneric Task A-40 is_ sui sivided into two phases.
cperations at operattrig plants, including design and procedural measur t at prevent or mitigate the con- Phase I includes a numbeQ: f subtasks related to the response of structures, syit$ ms, and components to sequences of a hea . ad handling accident and has earthquakes. These subtaq include studies on: (1) prepared a draft rep rt containing the NRC staff's quantifying conservatisrris In seismic design, (2) resolution of this issn including revised criteria and other recommendatims. This report is expected to be electro plastic seismic apidysis methods, (3) site-specific response spectraf (4) nonlinear structural ,
issued for public comrdentin January 1980. The report dyr.amic analysis procedu4, and (5) soil structure in-will provide the badd for revisions to the Standard ,
teraction. These studies ja e performed under NRC-Riview Plan (SRP) ay Regulatory Guides, if needed, sponsored contracts and $ were completed by Oc-that can be used in f ature reviews of new plants and l tober 1979. Review of tic' results of these studies is ;
will provida the basis'fbr implementing additional re- underway. The results w1 support the effort on seis-quirements and procidhres,in operating plants. ;
Although Task A mic reevaluation of opein'ing plants, particularly in ;
i will result in generic criteria, the area of site-specific ddinition of seismic input. As j
implementation of e criteria will be dependent on of January 1,1980, Phas W was scheduled to be com-plant design characte f ics and thespecific procedures pleted in February 1980mith the issuance of recom- 8 in effect at each parti ar plant, and will consequent- men,iations for changes nj the Standard Review Plan ly require a plant by lhnt review. and Regulatory Guides it those seismic design areas k related to response of stn etures, systems, and com- e 1 ponents to seismieully inneed ewnts. -
Seismic Design Criteria Phase II of Task As includes several subtasks l i]5 related to numerical mot ing of earthquake, motion.
NRC regulations rektlfre that nuclear power plant at the source, analysis of ar source ground motion, structures, systems anc components important to safe. and attenuation bf high@equency ground motiori.
ty be designed to wi Mstand the effects of natural Studies under th'ese subta being conducted by NRC contractors are scheduled @br completion by'the erid of5 phsnomena such as cujthquakes. Detailed require. 1980. Review and imple. $mtation of the results of mints and guidance r -_garding the seismic design of I nuclear plants is providellin the NRC regulations and these studies in terms of re +mmended revisions to th'e {
in Regulatory Guides. yowever, there are a number of Standard Review Plan a W Regulatory Guides are "
scheduled for March 1981. d {
plants with constructibn permits and operating W i lictnses issued before :h'e NRC's current regulations * - ' '
- g and regulatory guida de were in place. For this q rcason, rereviews of ;hb seismic design of various Pipe Cracks at Boiling Water Reactors j
plants are being undertden (principally as part of the
- Commission's Systemati n Evaluation Program) to {
assure that these plants E not present an undue risk to Pipe cracking has occurred in the heat affected i tha public. -
zones of welds in primary system piping in BWRs since The NRC staff is con:fheting Ceneric Task A-40 as the mid 1960s. These cracks have occurred mainly in ptrt of the NRC Progrim for Resolution of Ceneric Type 304 stainless steel that is being used in most l Issues. Task A 40 is a ch pendium of short-term ef- i forts to support the reev a ation of the seismic design operating BWRs. The major problem is recognized to o be intergranular stress corrosion cracking (ICSCC) of ;
of operating reactors, and a support licensing activity austenitic stainless steel components that h' ave been ;
in general. The objectiveldf the task is, in part, to in- made susceptible to this failure mode by being "sen- t vestigate selected areas onthe seismic design sequence sitized," either by welding or by post-weld heat treat-to determine the conservdtfsm for all types of sites, to ment. Although the likelihood is extremely low that '
investigate alternative app % aches to part of the design l
ICSCC-induced cracks will propagate far enough to f sequence, and to estimatquantitatively the overall create a significant hazard to the public, the occur-ccnservatism of the desig Esequence. In this manner I th2 program will aid the GC staff in performing its rence of such cracks is undesirable and measures to ,
minimize ICSCC in SWR piping systems are indicated i reviews of the seismic desg of operating reactors. to improve overall plant reliability.
I
... l
83
" Safe ends" (short transition pieces between vessel nozzles and the piping) that have been highly sensitiz-ed by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IG.
SCC in the late 1960s. Because they were susceptible 3 to cracking, the Atomic Energy Commission took the 7 position in 1969 that furnace-sensitized safe ends in older plants should be removed or clad with a protec- N D
tive material, and there are only a few BWRs that still y,A[g[ '
have furnace sensitized safe ends in use. Most of these, moreover, are in small diameter lines and are sub- C ,
fected to augmented inservice inspection. .
g <
Earlier reporte! cracks (prior to 1975) occurred primarily in 4-inch diameter recirculation loop bypass i lines and in 10-inch diameter core spray lines. More recently cracks were discovered in recirculation riser
- piping (12- to 14-inch) in all foreign plants. All these crack locations are part of the reactor primary system. ,,,,,,,,,,,,,,,,,
Cracking is most often detected during inservice in- came w ,ooccu u so N r spection using ultrasonic testing techniques. Some pip-ing cracks have been discovered as a result of small X
I.
primary coolant leaks. .
In response to these occurrences of BWR primary a ciacuunoa aisen pbhO -
MM system cracking, a number of remedial actions were " " ' " " ' ' " "
- 1L undertaken by the NRC. These actions included: a'caCvunom ourter 7 ,
c yy atCtacubAttCM INLET J
- Issuance of Regulatory Guide 1.44 on " Control of m3C,,a,73,,7,,","a'gyt IIN D .
the Use of Sensitized Stainless Steel." '
- Issuance of Regulatory Guide 1.45 on " Reactor ~
J: -
.\A ,
~
Coolant Pressure Boundary Leakage Detection MDj ;
Systems."
i
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~
- Closely following the incidence of cracking in . u BWRs, including foreign experience.
- Encouraging replacement of furnace sensitized 5"*"*"8"U" '"'* j ! ,
g .
safe ends. : 'fm5.
- Requiring augmented inservice inspection oflines evenu v wve
\ ~
having less corrosion resistant stainless steel, 7 w?, i .
especially those that have a high potential for W k-- . " [. s ~Wd ,/ht- .
~
cracking (service sensitive lines). ' '
- Requiring upgrading of leakage detection Cracks *PPaared at one facility la the so caHed " safe ends' he- -
systems. L**ca P iPas and vessel noules. This diagram shows the general location of the safu-and area (above), and detail of the cracising area (helow).
More recently pipe cracking and furnace-sensitized ,
safe end cracking have been reported in larger (24-inch diameter) lines in a CE-designed BWR in
- Cermany' with over 10 years of service. Because the
- safe ends in that facility had been furnace-sensitized .m. r su . ;
during fabrication, ICSCC was suspected. As a result ,
8 '""
of concerns regarding these furnace sensidzed safe ( 1 ends, a safe end was removed and subjected to destrue-l tive examination. During laboratory examination of " ""*" " i the removed safe end, including a small section of at- amenw eas to:hed pipe, cracks were discovered at various loca- '"^""***"'"
tions in the safe end and in the weld heat affected zone
- ^"*
- 1#
of the pipe. The cracks in the pipe weld area were very (" VEE ^**" .. v2."Mo"#Re's
'*^'""'~'
shallow with the maximum depth less than 5 mm (about 1/8 inch) in a wall thickness of abe,ut 1.5 inches. Cracking in the furnace-sensitized safe end, 'aae # oum also having a wall thickness of about 1.5 inches, was "vE '
"vE
84 '
I somewhat deeper. The German experience was the plants. The Study' Croup published its report first known occurrence of ICSCC in pipes as large as (NUREC-75/067) in October 1975, containing recom.
24 inches in diameter. mendations to reduce the incidence of ICSCC in sen-In June 1978, a through-wall crack was discovered sitized stainless steel piping. Following staff review of the Study Croup's recommendations, the staff issued in an Inconel recirculation riser safe end (10-inch diameter) at the Duane Arnold facility. The crack has an implementation document (NUREC-0313) which been attributed to ICSCC, although the material in established staff positions consistent with the recom-this Instance is different from the Type 304 stainless mendations of the Study Group.
steel that has been historically found to be susceptible As a result of the more recent incidents, the NRC re-to ICSCC. Prior to safe end removal, ultrasonic ex- established a second Pipe Crack Study Croup on Sep-amination showed several indications of possible tember 14, 1978. The new Study Group specifically cracks. Following their removal, cracking was addressed the following issues:
discovered in n!! eight safe ends. The cracking ap-
- The significance of the cracks discovered in large peared to have originated in a tight crevice between diameter pipes relative to the conclusions and the inside wall of the safe end and the internal thermal recommendations set forth in the referenced
- sleeve attachment. Such crevices are know to enhance report and its implementation document, ICSCC. Differences in materials, geometry, stress NUREC-0313.
levels, and crevices appear to make the problem at
- Resolution of concerns raised over the ability of Duane Arnold unique to a particular type of recircula- ultrasonic techniques to detect cracks in tion riser safe end (Type I). As a result of this event, austenitic stainless steel.
ultrasonic examination of the other Type I safe ends in
- The significance of the cracks found in large U.S. BWRs (i.e., at the Brunswick 1 and 2 facility) was diameter sensitized safe ends, and any recommen .
conducted. No significant indications of possible dations regarding the current NRC program for
.1 cracks were found in Unit 2 and one indication was dealing with this matter.
Identified at Unit 1. Although this latter indication
- The potential for stress corrosion cracking in was relatively minor and too small to be reportable. PWRs. -
pursuant to the NRC Regulations, periodic reevalua.
- The significance of the safe end cracking at tion of the Unit was deemed necessary. This ultrasonic Duane Arno'd relative to similar material and' indication at Brunswick Unit I was remeasured and design aspects at other facilities, reevaluated in the presence of NRC ultrasonic testing The new Study Group completed its evaluation in consultants at another plant shutdown in January February 1979 and issued a report, " Investigation and 1979. It was concluded that: (1) there 13 no apparent Evaluation of Stress-Corrosion Cracking in Piping of change of this indication between inspections, and (2) Iight Water Reactor Plants" (NUREC-0531). The although the existence of a very smalllocalized am of new Study Group not only reaffirmed the conclusions cracking cannot positively be ruled out, the most likely and recommendations reached by the previous group cause of this indication is irregularities at the weld-to- in NUREC-75/067, but also presented some new ideas base metal interface of the first bead weld at the ther- to reduce the potential for ICSCC and addressed the mal sleeve to safe end weld. This indication will be subject of ICSCC in safe ends. On March 13, 1979,.
reexamined during the next refueling outage. NRC issued a Notice in the Federal Register soliciting
- Ceneral Electric (the reactor vendor) has been asked public comments on NUREC-0531. After expiration of to provide an in-depth report on the significance of re- the public a:omment period and re, view of the Study cent events, including current inspection, repair, and Croups conclusions and recommendations, the staff renlacement programs. They were also asked to ad- Initiated Task A-42. The work to be performed under dress any new safety concerns related to the occur- Task A-42 was defined at that time as the' development rence of cracking in large main recirculation piping. of an update to the implementation document, Based on information presented by Ceneral Electric to NUREC-0313, to incorporate the new Study Group's date and on extensive staff evaluation, the staff con- e nelusi ns and recommendations and public com-cluded that the recent occurrences do not constitute a ments received on NUREC-0531.
basis for immediate concern about plant safety, nor re- Revision 1 to NUREC-0313 was issued in October quire any new immediate actions by licensees. 1979, and public comments have been solicited on the report. Revision 1 seu forth the NRC staff's revised Based on the earlier incidents of pipe cracking dis- guidel!nes for reducing the ICSCC susceptibility of j cussed above, the NRC formed a Pipe Crack Study BWR piping. The guidelines describe a number of Croup to: (a) investigate the cause of cracks, (b) make preventive and corrective measures acceptable to the interim recommendations for operating plants, and (c) NRC, including guidelines for: (1) corrosion resistant i recommend corrective actions to be taken for future materials for installation in BWR piping, (2) methods
' {
i 85 '
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I of testing,(3) processing techniques,(4) augmentedin. -
service inspection, and (5) leak detection. The report .*
3\3 i !i also included recommendations for developmental I.'w- .<
,? .. h- W,h
/ '
work to provide future improvements in limitint the Q g . QW,M
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extent of ICSCC or detecting it when it occurs. %f- 0 t, '
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I Contain ent Emergency Sump Reliability ,
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Following stulated loss-of coolant accident, "%%f,Q., % g.g :;&"~)k]&,g, T r5
.j
[ 1.e., a break in i reactor coolant system piping, the water flowing fr the break would be collected in
,j M;f;
.-m..: y -", u. r:'.[ :.1 the emergency su t the low point in the contain- gig (',[g.(9/g,
.,..<., ?:.p:g4 -
., j.; .
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ment. This water w id later be recirculated through
.6.M$GYJ'ij.. ,c1 y)S.i3p. ,,g1
, j J..s';r:., m )9 M 1 i
'.' m;;;gw #?.tW!
the reactor system b 4the emergency core cooling -
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pumps to maintain co Tooling. This water would also be circulated thro the containment spray 9" N. .Sj%;e'.hMyW jN$,%Q M. hMSMi[d.
system to remove heat an ion products from the .
h^,'. MQN ! $ E [l'hiY l containment. Loss of the ab 'ty to draw water from the emergency sump could herefore disable the 7- '
f s.t C 6 M i 1
emergency core cooling anc jont;tinment spray ' I i"2 M '
{
i systems. The consequences of the ulting inability to cool the reactor core or the con *at ent atmosphere h
could be melting of the core and/or ss of integrity in 4
the containment. [< g y.j.1{ "i,dg~- g~g y Q~is N
.' One potential way the ability to d water from 6~ J'i 3 q p N" y$@($ o
, h d ihh, a
a I
the emergency sump can be lost is from lockage by ' L.e.C._ %.
- j [k .-i 4 h,N )" '
debris. A principal source of such debris ild be the
'ty ?- 'W ',
j thermal insulation normally installed on t reactor coolant system piping. In the event of a pipi break, %gyj. -'
f ."'.M,,c.
?.7"h?l....ph 3 j ,,id. b\ A ' -
,%%.3 A
the subsequent violent release of the high , mssure t t water in the reactor coolant system could rip f the
.g crig $~ h,$ p,$*F ,
G. ~ <- i r -
insulation in the area of the break. The loose i la- P
- g. ( D *y,.
l tion material could then be swept into the sump nd M5Y
.f 3
.y'$ \y . , .% . 4:-
i block it. ,,,,,,,,,," .
, A Task Action Plan was under development t t
March 1979 when the Three Mile Island Unit 2 acci- gh;4 ..M.f.QjeM' \ A .gi, dent disrupted work on it. As of January 1,1980, the
- Task Action Plan was nearing completion. Nonethe. tMf-Q'WF Igy.7.$, [;g .M *~te..*j .k '"
g t
I less, several technical studies related to sump reliabill-ty which were already underway will either be incor- .c
"[Q9'd.4;, T d% h M .?
k q porated into Task A-43 or will provide input into l 't- *Q. '
Task A-43 elIort.s.
A study program investigating PWR vortex
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I technology has been completed by the Iowa Institute of Hydraulic Research and a technical report issued. A 9th !#. g N i fi".i . h.d g;L;g?,,?,@ a, g-j
)
summary report of NRC experience with containment $ WS. g sump testing is being prepared. This summary will be issued as a NUREC report in 1980. Based on the Iowa %';f)
- ,, F
. gi. N.4'(,.c @, p f ig(
9 g g T/
, g
- 7 . 9 .'6ys'g'M.,. ;f study program and the review of tests, NRR expects to- ~'
WIP# .
draft interim positions on sump design guidelines and preoperational test requirements in early 1980. R' WM K _ud M .
Criteria for the evaluation of operating containment A sumps will be fermulated at about the same time. NitC gaff memhen bicd to the North Anna Power Station Umt 1 in \ irginia to hiduct evalustions of the emcrency recir.
Finally, a program is being sponsored by the cui. tion .omp p.,, g, . ,w on cencrie goue, r g 4 a 73, Department of Energy, in cooperation with NRC, to phota at t.he top dio-s xt personnei lookins down inta the aid in resolving this issue as part of their Light Water i Safety Research Program. This is an experimental pro- a"j';[,/f,"d",'d",.
biocta, or prpe iyp etc. .r""*g,, "Q",y,"Cr*H"""Q l
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EXHIBIT 2 9
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705CO' LONG ISLAND LIGHTING COM PANY
'""r aaf,,g,w SHOREHAM NUCLEAR POW'ER STATION P.C. SCX 618. NORTM COUNTRY RO AD . WACf NG RIVER. N.Y.11792 May 15 1981 h'g5 :j$hud DE" l f*'id]li{""h SNRC-566. N .
A.
- REEI.STBEHINDTHEEE 5fAT ,
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation f_D *~'.4P f O IS$lg l M '
U.S. Nuclear Regulatory Commission Washington, D.C. 20555
'"Ag 7 -
3 , ., 7; Shoreham Nuclear Power Station - Unit 1 ' ^
Decket No. 50-322 _
C D
Dear Mr. Denton:
Forwarded herewith are six (.6) copies of LILCO's responses to the Safety Evaluation Report (SER) Outstanding Issues listed in Attachment 1. Responses to the remaining SER Outstanding Issues, listed in Attachment 2 are scheduled to be submitted on May 27, 1981, unless otherwise noted. -
Please note that our response to Outstanding Issue Number 8 .
" Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment" has been forwarded to you under separate
cover via letter SNRC-561, dated May 15, 1981.
- Very ly yours, ' "
nm - '
I
' J. P. Novarro '
- Project Manager V .
Shoreham Nuclear Power Station ;
RWG:ing \ l c .
y 0 E.hclosures oY t- ..
r cc:
.J.a. Eiggins
.a s %g.
.. i 1310 s 21 o y[,h 4
6 9-SNRC-566 ATTACEMENT 1 - SER OUTSTANDING ISSUES Number Issue 3 Piping Vibration Test Program - Small Bore Piping / Instrument Lines 5 LOCA Loadings on Reactor Vessel Supports -
and Internals -
13 NUREG-0619, Feedwater No:cle and Control Rod Return Line Cracking 14 Jet Pump Hold-down Beam 18 NURZG-0313, Rev. 1, Technical Report on Material Selection and Processing
" Guidelines for BWR Coolant Pressure Soundary Piping
- 22
- Appendix G - Impact Testing 36 Containment Purge System ,
38
- Fracture Prevention of Containment Pressure * **
Boundary .
44 Level Measurement Errors .
46 ', OIE Bulletin 79-27 .
50 Low and/or Degraded Grid Voltage Condition.
51
- Fracture Toughness of Steam Line and .
Feedwater Line Materials ,
56 Financial Qualifications (Supplement to' SNRC 561 dated April 30, 1981, issue 561 .
- BOP portion of response only
l/' -
t .
SER OPEN ITEM #13 NUREG-0619 "FEEDWATER NOZZLE AND f CONTROL ROD RETURN LINE C2ACKING"
, .RISPONSE:
Control Rod Drive Return Line No::le Crackine C
Postmodification CRD-System Performance Test '
~ The CRD System will be tested to prove satisfactory system operations return flow capability will be demonstrated to be equal to c= in excess of the flow required to satisfy the base-case conditions and two-CRD-pump operation, if concurrent operation of two CRD pumps are required to achieve this flow.
CRD Hydraulic System Maintenance Procedures For Shoreham, all lines are constructed of stainless steel and do not require procedural modifications for flushing the exhaust-water header and cleaning the filters in the insert and exhaust
- lines.
Triple Sleeve Sparger Design
- Tha welded thermal sleeve fee'dwater sparger has been replaced with an improved interference fit sparger,also called the triple sleeve sparger, for the Shoreham Nuclear Power Station. .
Clad Removal
- Fccdwater no==le cladding will not be used for the Shoreham: Nuclear Bower-Station. -
Syntem Modifications I
(n) Low Flow Feedwater Controller - the Shoreham Nuclear Power Station low flow feedwater control system consists of two low flow control valves in parallel with split range control. The split range control en' ables control to below 0.1% of rated flow. The basic. purpose of utili=ing a dual element controg scheme is to limit feedsater temperature fluctu-ations to within 50 F peak-to-peak on a continuous basis. Feedwater temperature variations will be monitored during start-up to evaluate the need for additional controls. If necessary, administrative con-trols and/or procedures will be developed to guide the operators during low flow operation to further minimize thermal cycling.
(b) Reactor Water Cleanup System (RWCU) - RWCU discharge currently has l
l the capability to deliver ficw to all feedwater no::les at Shoreham.
e
s ,
l 2
Operatine P;ocedure Mcdifications
~
, The following procedural modifications will be evaluated for Shoreham Plant operation:
(c) RWCU flow will be directed to all feedwater nozzles at maximum flow rate and exit temperature during all low flow conditions prior -
to turbine loading. -
(b) Low load, reduced pressure synchronization of the turbine will be evaluated during start-up.
(c) Maximum feedwater heating will be initiated as soon as possible ,
by isolating all feedwater heaters except the highest stage of feedwater heating.
(d) During start-ups and shutdowns, the feedwater control system will be operated to maintain low flow control sufficient to eliminate on-off feedwater operation and with sufficient controllability to preclude greater than 250F peak-to-peak mixture temperature varia-tions during steady demand.
- 00) In general, plant operating procedures will be modified to minimize-subcooling., particularly at high feedwater flow rates.. ,
Pro-Service Inspection . .
(a) Performance of PT has been accomplished for each nozzle prior to the installation of the sparger. .
(b) Performance of a baseline UT will be accomplished for.each nozzle after the installation of the sparger. Test results will be .
retained with the plant's permanent records.
In-Service Inspection ,
Ca) At every second scheduled refueling outage, an external. UT will be performed of all feedwater nozzle safe ends, bores and inside blend radii. If nozzle cracks are indicated, the spar.ger,will be. removed and a PT of the nozzle bore and blend radii will be. performed; and repairs will be made as required.
At every fourth scheduled refueling outage, a visual inspection of s
(b) all spargers will be conducted. .
l (c) At every ninth scheduled refueling outage (or after 135 startup/
l shutdown cycles)'a PT examination will be performed. The PT will i include removal of a sparger from one nozzle followed by flapper wheel grinding and PT examination of both the nozzle of the re-moved sparger and the accessible portions of the other nozzles.
o
, . , - - .- c _ ,-- , ,-,
Shoreham Outstandinc SER Issue #18 NUREG 0313
" Technical Report on Material Selection" l and Processing Guidelines for SWR Coolant !
Pressure Boundarv Picine j i
Both NRC (Pipe Cracking Study Group) and industrial study groups j such as GE and EPRI " Owner's Groups" have been investigating the . .
phenomena of BWR str'ess corrosion cracking for a number of years.
One common conclusion that has been drawn from these studies is that ;
the stress corrosion cracking issue is not.a safety issue. The '
conclur. ion that austenitic stainless steel will " leak before break- ,
ing" is substantiated and supported by all the above groups. In ,
addition, no incident to date has led to a major release of radio- ;
activity to the environment, and the relatively small cracks have t been identified either by leak detection or ISI. Therefore, we r believe the stress corrosion cracking issue is one of reliability, not safety, and sh uld be addressed accordingly. Although the i Shoreham plant's initial design did contain many of the " service censitive" lines outlined in NUREG 0313, Shoreham has been cognizant of these industry problems and.has taken steps to mitigate future concerns about stress corrosion cracking.
. Accordingly, we believe.that the problem of stress corrosion. .
cracking has, to a large degree, been eliminated for Shoreham. In particular: (a) the recirculation bypass line has been elimin,ated, ,
(b) the core spray line and safe end materials have been changed ,
r and, (c) the CRD return line has been eliminated. Therefore, the only primary pressure boundary piping which has remained 304. .. l stainless steel.is the reactor recirculation system. A modification of the piping material for this system to an alternative: material, f was not practical in the time table for Shoreham. -
l In an effort to further reduce the potential for stress corrosion. ..
cracking in the recirculation system, sections of the shop fabri . .
cated piping were solution heat treated prior to field erection. .
These piping sections were received from GE in a non-heat treated .
condition. The heat treating should eliminat's all concerns about."
stress corrosion cracking for the shop welds that were. heat treated.
o For the field welds,LILCO did institute a number of measures whereby i
the welding techniques used were modified such that sensitization was minimized and the welding residual stresses introduced during i the welding process were reduced to the best extent practical. '
- These measures included
- (1) welding process control (i.e., pbeheat temperature and interpass temperature control), (2) grinding re-strictions, and (3) weld filler material ferrite control as required by Regulatory Guide 1.33. -
c- --m- -- -, , - - - -
o .- ,
a se
-2 .
4 MUREG-0313, Revision 1
- To address NUREG-0313 specifically for Shoreham, the ASME Code j l
Class 1 and Class 2 reactor cooling pressure boundary piping ;
mnets the guidelines stated in Part 2 except for portions of the l reactor recirculation system (331) as described above, and, stain- l loss steel to carbon steel transition welds between the (
Recirculation System and the Reactor Water Cleanup, Core Spray' >
and Residual Heat Removal System. As stated previously, only - .
f field welds and shop welds classified as " nonconforming" would fall under the requirements for Part 3 of NUREG-0313. !
Part 3: -
A. 1.0 ISI i The recirculation system welds and transition welds classified as " nonconforming" will be inspected in accordance with Part i III Bl.0 of NUREG-0313 to the extent practicable for Shoreham Nuclear Power Station. Attempts will be made to inspect all welds with the appropriate.ISI UT techniques, but physical ~
interference in some locations may preclude the inspection.
Although NUREG-0313 Revision 1 implies that recirculation '
riser lines and recirculation inlet lines at safe and curves .
should be considered " Service Sensitive", we would argue that this has not been documented for BWR 4 designs such as . . .
Shoreham. First, most of the riser sections for Shoreham have.
been solution heat treated and, therefore, IGSCC is not a ,
concern. For the recirculation system heater to riser welds..
these welds will be inspected as Nonconforming Lines, but .
non-Service Sensitive. For recirculation nozzles / thermal
- sleeve designs similar to Shoreham (i.e., Fitzpatrick), no
- evidence of cracking has been discovered. Therefore, until further IGSCC incidents are documented in these areas we believe it is appropriate to classify these areas as Noncon-
~
forming, but Non-Service Sensitive.
\ .
2.0 Leak Detection .
The Shoreham primary containment leakage detection system is in full compliance with Regulatory Guide 1.45. The technical :
specification limits for reactor coolant leakage will be in.
cludes as part of the Technical Specification submittal for Shoreham. We believe a change in limiting conditions for' leakage is not warranted since this issue is not a safety I concern and thus the present limiting conditions are acceptable.
4 i
G e = =
I If any cracks are detected, all spargers will be , removed and a complete examination'of all nozzles will be conducted. All no::le cracks will be evaluated and repaired as required.
(d) An on-line monitoring system for detecting leakage through degraded seals and/or cracks in thermal sleeve welds will be evaluated and
~
considered for Shcreham feedwater nozzle application after fuel load, in lieu of continued in-service inspection.
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1 SNPS-1 FSAR 5.2.3 General Material Considerations 5.2.3.1 Material Soecifications Table 5.2.3-1 lists the principal pressure retaining materials, and the appropriate material specifications for the RCPB components.
5.2.3.2 Compatibility with Reactor Coolant The materials of construction exposed to the reactor coolant con-
, sist of the following: .
- 1. Solution annealed austenitic stain 1,ess steeds (both ~
wrought and cast) Types 304, 304L, 316, and 316L. .
- 2. Nickel base alloys - Inconel 600 and.Inconel X750.
- 3. Carbon steel and low alloy steel.
- 4. Some 400 series martensitic stainless steel (all tempered at a minimum of 1100 F) .
- 5. Colmonoy and Stellite hardfacing materials. .
All of these mater'ials of construction are resistant to stress corrosion in the BWR ccolant. General corrosion on . all' materials, except carbon and low alloy steel, is negligible.
Conservative corrosion allowances are provided for all 'exo'p sed surfaces of carbon or low alloy steels. * '
Contaminants in the reactor coolant are controlled to veiy low limits by the reactor water quality specifications. , No detrimental effects will occur on any of the materials from
! allowable contaminant levels in the high purity reactor coolant.
Radiolytic products in a DWR have no adverse effects on the.
constructica materials. * * * '*
5.2.3.2.1 Use of Sensitized Stainless Steel . .
The use of severely sensitized or furnace. sensitized stainless steel is prohibited. However, exception is taken' to Regulatory
=
Guide 1.44 which effactively prohibits the use of ' welded type 304 l stainless steel by considering it the same as severely sensitized -
material and by placing limits on oxygen concentration which cannot be met by operating BWRs. ..
The stated objective of the Regulatory Guide is to control the application and processing of stainless steel to avoid severe l
sensitization "to M m%ish the numerous occurrences of stress
! corrosion cracking in sensitized stainless steel components' of nuclear reactors.u
9 SNPS-1 FSAR -
Welded 304 (and 316) stainless steel is placed in the same category by the guide as severely sensitized stainless steel. An intergranular corrosion test is required for' e'ach welding procedure to be used for welding of material with carbon content greater than 0.03 percent. Limits are imposed upon water quality which are inconsistent with Regulatory Guides 1.37 and 1.56. The use of welded 304 stainless steel is prohibited, primarily by imposing an oxygen limit of 0.10 ppm which is not obtainable in the BWR during normal operation.
Our position is that present controls administered by General Electric Company (GE) for processing and application of welded 304 stainless steel are entirely adequate, have been well proven by excellent operating experience, and therefore the requirements imposed by the Regulatory Guide in its present'. form are not warranted. .
That there have been instances of stress corrosion cracking (SCC) in welded 304 stainless steel in some early GE BWRs is indisputable. Also, there have been instances of SCC in furnace sensi.tized stainless steel. However, there have been no instances of SCC of 304 stainless steels which were processed with presently implemented controls.
5.2.3.2.2 Process Controls for 304 Stainless Steel Improvements in technology fostered by extensive GE-APED research .
and development since the early GE BWRs have resulted in imple-mentation of the following ma-Jor processing controls for 304,,,
stainless steel: .
- 1. Furnace sensitized components are pr'ohibited. . .,
- 2. Welding heat input is restricted to 110,000 joules /in. -
and a maximum interpass temperature
~
of 350 F is -
required.
- 3. Block welding is prohibited. , ,
- 4. Restrictions are placed on cold work. .
~
- 5. Fabrication and cleaning controls are- specified. to.'
minimize cont =% ants. .
- 6. Pickling of welded stainless steel is prohibited. -
The effectiveness of these controls has been well demonstrated, j by a
absence of a single stress corrosion cracking incident in normal" SWR service, in the 5 years since they were implemented.
(Note: " normal" is used to distinguish from abnormal service such as chloride intrusion.)
C; h10
6 EXHIBIT 4 SECTION 6.1 - 6.3 of NUREG-0531 e e e .
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6.0 PIPE CONFIGURATIONS AND STRESSES *
6.1 INTRODUCTION
There are approximately 17,000 (6.1) welds (a) .in stainless-steel piping in operating BWRs. There have been 133 IGSCC incidents (b) in BWR stainless-steel piping systems. The j j question arises: Why have less than 1% of the welds cracked while more than 99% of the f 1
welds are still perfonning satisfactorily? The variations in detailed material character-I istics and environmental conditions are probably major contributors to the 1% failure. 1
- However, the variability of stresses at welds may also help to explain why less than 1% of the welds have cracked.
Stresses at welds are caused t,y fabrication processes (e.g., weldirg) and by operatibn * ~ ~
loads (e.g., internal pressure). Stresses from both fabrication processes and operation loads are dependent upon the pipe configuration. Accordingly, this chapter will outline j first significant aspects of pipe configurations and then discuss fabrication and operating i
stresses and how those stresses can be influenced by configurations.
I i 6.2 PIPE CONFIGURATIONS 6.2.1 Pioing Systems .
f Tabla 6.1 lists typical BWR piping systems that are involved in the operation and/or shfe shutdown of the reactor. Figure 6.1 shows the basic configuration of BWR recirculation ,
j TABLE 6.1. List of Typical Piping Systerns Involved in the Operation and/or Safe Shutdown of BWRs(a) .
Identification . ..
Number Function I '
1 Reacor Recirculation 2 Main Steam 3 Feedwater ,
e * * * * *
- 4 Reactor Water Ceanup 5 Reactor Core Isolation Cooling '. '
6 Core Spray .
8 Containment Spray - -
9 Reactor Head Spray 10 Standby Liquid Control 11 High. Pressure Coolant injection 12 Low. Pressure Coolant injecion .
l (a)This list is taken from the FSAR of Peach Bottom Units 2 and 3.
(a) The term " welds" is used in this chapter to refer to girth-butt welds.
(b) This does not include IGSCC .that occurred in furnace-sensitized stainless-steel safe ends.
6.1 1
l
.'
- piping systems. While it does not show the various crnnecticns (e.g., rssidual heat rePoval) to the system or the support /r: straining devices, it is sufficient.fer th) pr:s;nt purpass because it shows that the recirculation piping system is a c'omplex assembly of piping products consisting of several sizes of straight pipe, curved pipe and branch connections.
Figures 6.2 and 6.3 are additional illustrations of the complex nature of BWR piping systems.
2 etAc7ca Petssuat
. vtssa siaccicas,
} sHlILD v
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0
, -- -alsets c, e- _. r=:
- A PLMP
- . py ,, OlsCH48GI sucnow - ""
MAIN :r SYPAss i
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FIGURE 6.1. Basic Configuration of BWR .
- Recirculation System
[ Source: (6.2), p. 2.2)
- 6.2.2 Welds Weld configurations are of particular significance to stresses in IGSCC. Pipe, fittings '
and valves are not precision-made products; the relatively large tolerances on diameters and -
wall thickness cause problems in achieving an adequate fit between two pipes (or pipe-to-
- fitting or pipe-to-valve) for butt welding. It is connon practice to counterbore ends of ,
pipe, fittings and valves to a standard 'C-dimension " as illustrated in Figure 6.4*. The "
locally-reduced wall thickness in the weld region may produce a significant increase in opera-tion stresses as compared to such stresses remote from the weld region. The counterbore .
depth (d in Figure 6.4) is not a controlled dimension and usually varies with azimuth (around .
the circumference) location of a given pipe end. This may intreduce variations in residual -
welding stresses witn azimuth location.
- There are many welds in piping systens. As typical examples, the Browns Ferry 1 recirculation piping system contains thirty 4-in., sixty 12-in., nine 22-in, and thirty-three 28-in, welds: a total of 132 welds. The Browns Ferry 1 core spray piping systens','up l to the first isolation valves, contain six 10-in. and forty-eight 12-in welds.
6.2
g D
L I .
v i .
b 11
- s I
r FIGURE 6.2. Dresden.2 Recirculatfort Bypass Lines - Loops A and 8
[ Source: (6.2), p. 2.8]
P f
maimersus w
- s L.,
FIGURh 6.3. Dresden-2 Core Spray Loop A (North) (Source: (6.21, p. 2.3)
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- C IL j i l*lL3N"
- Cell 95" us Missueu WL3 IOutDe TWO PtFL Dies hEa
- \L1T CE0ll ECT19s USlaMLT vaattiene ClaCINIFUturlaLLOCAII0le f:=~ ' A:-,
ifIntes OG .- f i . .
- p??c5 b :.h.*lY+).S W'Y p ,, 2 ::p .
e i-e Otput3DsIAftvt utLS 104D1fintar.
De vatyt Asse MPt FIGURE 6.4. Configuration of Welds (Girth Butt Welds) 6.3 1
- . i
- .' O. 6.2.3 Oth c Detailed Configurations asm Piping systems contain curved pips, branch connections (including crosscs in the recirculation loop Figure 6.1) and other structurally complex shapes. Stresses in these products are different from those in straight pipe with the same operation loads. Further, welding residual stresses at welds between pipe and a valve, for example, may be different from stresses in a pipe-to-pipe weld. (As discussed in 6.3, some data are available on '
residual stresses in pipe-to-pipe welds.)
The Duane Arnold (see Chapter 7) recirculation inlet nozzle represents a type of detailed configuration of particular interest to this report. This type of configuration, in addition to the crevice problem, may involve relatively high stresses that are due to operation loadings such as pressure, pressure differential across the interior pipe, loads -
(possibly including vibration) transmitted to the attachment weld by the interior piping and differential thermal expansion between the interior pipe and the safe end or vessel nozzle.
This type of detailed configuration is also encountered at the connection of interior piping ,
to the core spray nozzles of the reactor vessel. The attachment weld 'will create residual ,
stress and, if the pressure boundary material is susceptible, sensitization may occur.
6.3 FABRICATION STRESSES Stresses (often called residual stresses) can arise because of forming processes (e.g.,
bending of pipe), welding, rough grinding or machining and forces required t'o obtain align-ment for making the final weld (field closure weld) that unite the piping subassemblies into a piping system.I*I If the piping is annealed after these fabrication processes, the fabri-cation stresses are essentially eliminated. However, annealing after the field closure weld is usually not practical.
The mest significant fabrication stresses in connection with IGSCC are probably due to welding and to rough grinding without annealing. - -
6.3.1 Weld Residual Stresses .
Welding affects IGSCC in two ways. First, and perhaps most important, the high *
temperatures that occur during welding may sensitize the material adjacent to the weld. .
Second, welding produces residual stresses in the weld region; these residual stresses. -
either by thanselves or in combination with operation stresses, provide the stress ingredient of IGSCC. . . .
Test data (6.3, 6.4) are available on residual s' tresses at welds 4-in.,10'in, - and' 26-in.
Schedule 80 Type 304 stainless-steel pipe-to-pipe welds. Test welds were made using pro- .
cedures (e.g., number of passes, heat input) that were representative of M.andard nuclear- .
~ '
industry practices. Shack, Ellingston and Pahis include some residual stress data on welds -
, in piping that was removed from BWR service (6.4). '
The test data indicate that residual stresses in the region of welds vary in complex ways in all spatial directions; i.e., around the circumference (azimuthal location), along the axis of the pipe as a function of distance from the weld, and as a function of location (a)Insomepipingsystems,intentionalforcesatthefieldclosureweldintroduce" cold spring." In the SWR piping systems considered here, this is not done. 1 6.4 1
y -
-- ' ~~ -
, - . T - . . . . _ . . - . - -_. .-
l Ie- through the wall of the pip 3. From the standpoint of IGSCC, the most significant residual
! stress is probably the tensilo axial-direction stress on th2 inside surface within an axial distance of about 1/4 in. from the weld. The reasons for this are:
1
- Tensile stresses are more important than compressive stresses in IGSCC.
l
- e Axial-direction stresses will cause cracks which folicw along the edge of the weld l and stay in the sensitized metal zone, in contrast to hoop stresses, which will l
, 3 cause axial cracks that tend to stop at the weld or upon leaving the sensitized l metal zone.
l e Cracks are most apt to begin on the inside surface.
!
- The 1/4-in axial zone on each side of the weld appears to be the most severely weld-sensi tized.
- I Giannuzzi (6.3) conducted tests on one 4-in., one 10-in. and two 26-in. Schedule 80
! welds and found that the maximum tensile, axial-direction stress on the inside surface heat-I affected zone (sensitized zone) could be ranked as follcws: ,
j highest: 4-in. Sch. 80 intermediate: 10-in. Sch. 80 i lowest: 26-in. Sch. 80.
The stress in the 4-in, weld was about 20 to 25 KS! higher than in the 26-in weld.
Shack, Ellingston and Pahis (6.4) conducted tests on three 4-in., one 10-in., one
! 26-in., Schedule 80 welds and found that the maximum tensile, axial-direction stress on the l inside surface in the region 2 to 3 mm from the weld fusion line were:
4-in, welds: 38, 46, and 51 KSI for the three tests 10-in weld: 60 KS!
26-in weld: 28 KSI The 10-in weld was removed from a piping system in a BWR. .
These conclusions suggest that welding residual stresses in 26-in.' Schedule 80 p,ipe are less than in 4-in. Schedule 80 pipe. This is perhaps related more to the wall thickness of the pipe and number of passes (7 passes for 4-in., 30 passes for 26-in.):than to the si2e or I diameter of the pipe.
Residual stresses are quite high, even in the 26-in. pipe tests. However., the , , . ,
differences between residual stresses in 4-in. and 25-in. pipe may be part of the reason why IGSCC is more frequent in small pipe than in large pipe.
Discussions of available test data (6.3, 6.4) lack rational explanations for the compl,ex distributions and magnitudes of measured stresses. Efforts have..been made to calculate residual stresses using finite element modeling of the structure and materials character-istics (6.5). However, at this time, the quantitative generality of conclusions orawn from the relatively scanty data is questionable. Also, tests were run on pipe-to-pipe welds, d
6.5 1.
thereas many welds in piping systems are pipe-to-fitting (a) or pipe-to-valva welds. It is not apparent that rr.sidual str:.sses in such welds will ba equivalent to r;sidual stresscs in pipe-to-pipe welds. -
6.3.2 Rouch Grinding or Machinino Stresses Preparation of welding ends involve machining of a counterbore to obtain good alignment of the inside diameters at the pipe-to-pipe juncture; this alignment is needed to produce a weld with a smooth root and a weld tnat can be inspected. The counterbaring is usually done by relatively rough machining; this can lead to high ' residual stresses on the counterbore surface. However, according to Shack Ellingston and Pahis (6.4), welding tends to anneal-out these surface stresses in the weld-sensitized region close to the weld; hence machining or grinding before welding is probably not a major contributor to IGSCC. However, studies (6.3, 6.4) indicate that rough grinding after welding (as might be needM to smooth out an 4
irregular weld root contour) can lead to high (up to 100 KSI) surface residual stresses.
While these high, surface residual stresses extend only a few mils into the wall, they may initiate IGSCC. -
i
! .6.4 OPERATION STRESSES Operation stresses are those caused by operation loads such as' Sternal pressure;
, weight of pipe, contained fluid and insulation; restraint of overall thermal expansion of the piping system; temperature gradients through the wall; and dynamic loads such as safety relief valve thrust, earthquakes and vibration.
- For evaluation of IGSCC, those 5, tresses that are present over Ic6g periods of time (such
, as those produced by pressure on restraint of thermal expansion) are probably more significant
{ than those present over relatively sher,t periods of time (such as those produced by tempera-3 ! ture gradients or safety relief valve thrust). However, even though these stresses are of I
short duration, they may influence propagation of racks because of their cyclic nature. ,
6.4.1 Code Limits on Ooeration Stresses ., ,
i The magnitude of operation stresses is limited by the codes under which the piping ,,
systems are designed. For BWRs now in operation, the code used in design was ANSI B31.1.0, I " Power Piping." The ASME Boiler Code Section III, " Nuclear Power Plant Components," is the i
current code used for design of piping systems in nuclear power plants.
I Design codes for piping use a rather complex set of stress limits ,none of which , , ,
includes allowances for metal deterioration such as IGSCC. A nominal " corrosion allowance" may be included and some warnings may be given (e.g., graphitization of steels' at elevated temperatures), but, in general, the designer is expected to use materials that will not , '.
deteriorate in the anticipated environment. Also, desi7n codes for piping do not consider
- I f (a) Shack, Ellingston and Pahis (6.4'; include residual stress data on 4-in. pipe-to-elbow welds from piping that was removed from BWR service. The authors speculate that: "The relatively low (compared to pipe-to-pipe test welds) stresses seen in the straight-pipe-to-elbow weldments are most likely due to the different restraint imposed by the elbow geometry. The two elbow welds may also have been stress-relieved (annea, led);
however, metallurgical examination, while not conclusive, suggests that this is not the case."
4 6.6 s -
- . ....._m._
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 97 \q( -6 Y0 $
)
In the Matter of ) ;.. 3
) .
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 0.L.
)
(Shoreham Nuclear Power Station, )
Unit 1) )
)
CERTIFICATE OF SERVICE I hereby certify that copies of the following documents are to be served on the persons indicated by an asterisk (') by hand delivery on May 4, 1982, and to the remainder of the Service List by placing them in the mail, first class, postage prepaid, on May 4, 1982:
- 1. Testimony on behalf of Suffolk County on Suffolk County Contention 24 and SOC Contentions 19(c) and 19(d);
- 2. Testimony on behalf of Suffolk County on Suffolk County Contention 26;
- 3. Testimony on behalf of Suffolk County on Suffolk County Contention 28 (a) (i) and SOC Contention
- 7. A (1) ;
- 4. Testimony on behalf of Suffolk County on Suffolk County Contention 28 (a) (iii) and $0C Contention 7.A(3);
l l
L
- 5. Testimony on behalf of Suffolk County on Suffolk County Contention 31;
- 6. State of Qualifications of Robert Neil Anderson and Vera Barlit.
Lawrence Brenner, Esq.*/ Matthew J. Kelly, Esq.
Administrative Judge Staff Counsel Atomic Safety and Licensing Board New York State Public Service U.S. Nuclear Regulatory Commission / Commission Washington, D.C. 20555 3 Rockefeller Plaza
, Albany, New York 12223 Dr. James L. Carpenter _/
Administrative Judge Stephen B. Latham, Esq.*/
Atomic Safety and Licensing Board Twomey, Latham & Shea U.S. Nuclear Regulatory Commission 33 West Second Street Washington, D.C. 20555 P.O. Box 398 Dr. Peter A. Morris */
Administrative Judge Marc W. Goldsmith Atomic Safety and Licensing Board Energy Research Group, Inc.
U.S. Nuclear Regulatory Commission 400-I Totten Pond Road Washington, D.C. 20555 Waltham, Massachusetts 02154 Edward M. Barrett, Esq. Mr. Jeff Smith General Counsel Shoreham Nuclear Power Station Long Island Lighting Company P.O. Box 618 250 Old Court Road .
Wading River, New York 11792 Mineola, New York 11501 David H. Gilmartin, Esq.
Mr. Brian McCaffrey Suffolk County Attorney Long Island Lighting Company County Executive / Legislative Bldg.
175 East Old Country Road Veterans Memorial Highway Hicksville, New York 11801 -
Hauppauge, New York 11788 Ralph Shapiro, Esq. Hon. Peter Cohalan Cammer and Shapiro Suffolk County Executive 9 East 40th Street County Executive / Legislative Bldg.
New York, New York 10016 Veterans Memorial Highway Hauppauge, New York 11788 Howard L. Blau, Esq.
217 Newbridge Road Jay Dunkleberger, Esq.
Hicksville, New York 11801 New York State Energy Office
, Agency Building 2 W. Taylor Reveley III, Esq. j Empire State Plaza Hunton & Williams Albany, New York 12223-707 East Main Street
! P.O. Box 1535 Richmond, Virginia 23212 l
Ezra I. Bialik, Esq. Docketing and Service Section Assistant Attorney General Office of the Secretary Environmental Protection Bureau U.S. Nuclear Regulatory Commission New York State Department of Law Washington, D.C. 20555 2 World Trade Center New York, New York 10047 Atomic Safety and Licensing Board
/ Panel MHB Technical Associates U.S. Nuclear Regulatory Commission 1723 Hamilton Avenue, Suite K Washington, D.C. 20555 San Jose, California 95125
, Atomic Safety and Licensing Appeal Bernard M. Bordenick, Esq. p / Board David A. Repka, Esq. U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 J. Letsc e" ')
Marl)
KIRK PATRICK,L[OCKHART, HILL, CHRISTOPHER & PHILLIPS 1900 M Street, N.W.
8th Floor Washington, D.C. 20036 May 3, 1982 l
1