ML20024D872
| ML20024D872 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/21/1983 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Parris H TENNESSEE VALLEY AUTHORITY |
| References | |
| IEB-82-03, IEB-82-3, NUDOCS 8308080321 | |
| Download: ML20024D872 (5) | |
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July 21,1983 Docket No. 50-296 Mr. Hugh G. Parris Manager of Power Tennesse Valley Authority 500 A Chestnut Street, Tower II Chattanooga, Tennesse 37401
Dear Mr. Parris:
SUBJECT:
INSPECTIONS OF BWR STAINLESS STEEL PIPIN3 As you know, the Commission has recently been briefed on the results of inspections and tests conducted on BWR recirculation system and residual heat removal system piping. Based upon these discussions, the NRC has concluded that it is important to obtain additional information regarding the extent of intergranular stress corrosion cracking (IGSCC) at your facility as soon as practicable.
As a result of inspections conducted at 18 operating Boiling Water Reactors (BWRs) in conformance to recent IE Bulletins (IE Dulletin No. 82-03, Revision 1
" Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculatio1 S.vstem Piping at BWR Plants," and IE Bulletin No. 83-02 " Stress Corrosion Cracking in large-Diameter Stainless Steel Recirculation System Piping at BWR Plants"), extensive IGSCC in the primary system piping has been discovered.
These bulletins requested selected licensees to perform a number of actions regarding inspection and testing of pipe welds.
The Commission is considering action to accelerate the conduct of inspections by BWR licensees that have not yet begun inspections requested by the IE Bulletins.
Inspections conducted in response to these bulletins over the last nine months highlighted the IGSCC problem. Results of these and other inspections have revealed extensive cracking in large-diameter recirculation and residual heat removal system piping. For these plants, repairs, analysis and/or additional surveillance conditions were required. The NRC has concluded that other uninspected BWR-facilities may have similar IGSCC, which may be unacceptable for continued safe operation without inspection and repair of the affected pipes and additional surveillance requirements. For this reason, we are desirous of accelerating the inspections as much as possible.
Accordingly, pursuant to 10 CFR 50.54(f), you are requested to submit, signed under oath or affirmation, the information described below. This information should be provided as soon as possible, but in any event, it must be received by the NRC by August 4,1983.
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A justification for continued operation of your facility prior to completing the inspections tescribed by Attachment A in view of the increased evidence of cracking since the issuance of IE Bulletin 83-02.
2.
Identify any weld inspections which appear to satisfy the sensitivity for detection specified by IE Bulletin 83-02.
The information provided should include a list of these inspections, the dates of the inspections, the extent and results of those in'spections, and a description of the technique or equipment used.
If you have concluded that these previous inspections should influence the scope or schedule of the inspections described in Attachment A, please provide the basis for your conclusion.
Further, describe any other Lnique safety relcted feature, information or. action that would justify not accelerat-ing your current test and inspection schedule in accordance with IE Bulletin 83-02.
3.
Describe any special surveillance measures in effect or prcposed for primary system leakage in addition to the current Technical Specification requirements for your facility.
4.
Direct and indirect costs and impact, including efftets on other safety related activities, of conducting the inspections described in Attachment A:
(a) at a time which you would commit to conduct the inspections consistent with Chairman Palladino's suggestion to the staff and licensees that a realistic schedule for the inspections be develcpad "with the idea of accelerating the inspection as much as possible,"* and (b) at the time of your next scheduled refueling outage.**
5.
The direct and indirect ccsts and impact, including effects on other safety related activities, of suspending operation to 'litiate the inspection described in Attachment A within each of three possible times:
(a) 30 d'ays, (b) 60 days, and (c) 90 days from August 15, 1983.**
6.
A discussion of the availability of qualified inspection personnel i
to perform the inspection described in Attachment A at your facility for the various options in items 4 and 5, above, and the steps you have taken to obtain the services of. such personnel, i
Transcript of Commission's " Meeting with Industry -- BWR Owner's Group,"
July 15,1983, page 101 (See Attachment B).
- In providing the information requested in items 4 and 5, the bases for the cost information and the reasons for the effects on other safety related activities should be discussed.
In addition, the option of conducting the required inspections at the specified times separate from or as a part of a enfnolinn notaan. and the costs and benefits c f such an ootional aooroar.h should te discussed.
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-3 Upon receipt of the information requested above, the staff will arrange a meeting to discuss your submittal. The schedule for the meeting will be arranged through the assigned NRC Project Manager for your facility.
If you need to request an extension of time for the submittal of the required infonnation, such a request shall set forth the justification for the delay and include a proposed schedule for supplying this infonnation. Such a request shall be directed to the Director, Division of Licensing, NRR. Any such request must be submitted no later than July 28, 1983.
Sincerely.
ngginni cisd W -
paroll G= M Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosures:
1.
Attachent A. Guidelines for Conducting Inspections and Tests of Large Diameter Pipe Welds
- 2., Partial Transcript of the Commission's " Meeting 4
with Industry -- BWR Owner's Group," July 15, 1983, page 101 cc w/ enclosures:
See next page DIST: Docket File NRC PDR LPDR ORB #2 Reading OELD EJordan 1
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Mr. Hugh G. Parris Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2 & 3 cc:
H. S. Sanger, Jrl, Esquire U. S. Environmental Protection General Counsel Agency Tennessee Valley Authority Region IV Office 400 Commerce Avenue Regional Radiation Representative E 118 33C 345 Courtland Street Enoxville, Tennessee 37902 Atlanta, Georgia 30308 Mr. Ron Rogers Resident Inspector Tennessee Valley Authority U. S. Nuclear Regulatory Commission 400 Chestnut Street, Tower II Route 2, Box 311 Cnattanooga, Tennessee 37401 Athens, Alabama 35611 Mr. Charles R. Christopher Mr. Donald L. Williams, Jr.
Chairman, Limestone County Commission Tennessee Valley Authority P. O. Box 188 400 West Summit Hill Dr., W10395 Athens, Alabama 35611 Knoxville, Tennessee 37902 Ira L. flyers, M.D.
State Health Officer George Jones State Department of Public Health Tennessee Valley Authority State Office Building P. O. Box 2000 Montgomery, Alabama 36130 Decatur, Alabama 35602 h
Mr. H. N. Culver Mr. Oliver Havens 249A HED U.S. Nuclear. Regulatory Commission 4'.0 Cc.merce Avenue Reactor Training Center Tonnessee Valley Authority Osborne Office Center, Suite 200 Knoxville, Tennessee 37902 Chattanooga, Tennessee 37411 James P. O'Reilly Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303
ATTACHMENT A U
GUIDELINES FOR CONDUCTING INSPECTIONS AND TESTS OF LARGE DIAMETER PIPE WELDS Enclosed is IE Bulletin 83-02, " Stress Corrosion Cracking in Large-Diameter Stainless Steel Recirculation System Piping at BWR Plants," issued on March 4, 1983.
The following items reflect an expansion to the IE Bulletin.
1.
Conduct inspections and evaluations as requested in the IE Bulletin 83-02.
2.
In addition to the examinations required in the IE Bulletin 83-02, the licensec should conduct an ultrasonic examination of six or 100% of the welds (if there are fewer than six) in the ASME Code Class 1 portion 1
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of the residual beat removal system.
a.
If flaws indicative of cracking are found in the above 4
examination, additional inspection of the welds within the ASME Class 1 portion of the RHR system should be conducted in accordance with DdB 2430 of the ASME Code Section XI and appropriate action should be taken.
b.
The ultrasonic inspection and evaluation of the RHR piping welds shall be in accordance with those specified in IE Bulletin 83-02.
c.
Inspection reporting requirements shall be in accordance with those specified in IE Bulletin 83-02.
3.
The note on page 4 of IE Bulletin 83-02 is not applicable for the inspections described in items 1 and 2, above.
I
/
OMB Ns.:
3150-0096 a.
Expiration 0;to:
12/31/84 IEB 83-02 UNITED STATES NUCLEAR REGULATORY CDPWISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 4, 1983 IE BULLETIN NO. 83-02:
STRESS CORROSION CRACKING IN LARGE-DIAMETER STAINLESS STEEL RECIRCULATION SYSTEM PIPING AT BWR PLANTS Addresseq:
Those licensens of operating boiling water reactors (BWRs) identified in Table 1 for actio1.
All other licensees and holders cf construction paraits (cps) for information only.
Furr.ose:
t IE Bulletin 83-02 is issued to further inform all licansees and CP holders about the recent generic pipe cracking problems involving SWR plants and to require actions of those licensees listed in Table 1.
Escription cf Circumstanc_es,:
As a nsult of the axtensive intergranular stress cormsion cracking (IGSCC) found at Nine ? tile Point Unit 1, ths NRC issued IE Bulletin 82-03, Revision 1 fer action te nine BWR plants schaduled for refueling outages in late 1982 and early 1983.
Inspectient pursuant to IFB 82-03, Revision 1, and NUREG-0313, Ravision 1, have shown cracking of the main recirculation system piping in fivo of seven plants examined,to date. Table 2 presents a summary of affected plznts based on information available to date.
IEB 82-03 Rev.1 discusses the IGSCC problems experienced at Nine Mile Point Unit 1.
A brief description of tha' cracking problems at Browns Ferry Unit 2, Monticello and Hatch Unit 1 is prasented below.
At Browns Ferry Unit 2', the' inservice inspection (ISI)~was extended to include tha welds joining the jet pump piping sweepolets to the manifold of both A and B loops.
Unacceptable indications were found in the heat-affected zone of the l
manifold in the loops A and 8 sweapolet-to-manifold joint nearest the end caps.
All of the indications were interpreted to be cracks near the inside surface and wsre determined by UT to be about lis inches long (roughly parallel to the weld),
i and of about 20 percent depth through-wall.
As a result of further design annlysis, review of shop fabrication records, and supporting in situ metallography and ferrite determinations, the licensee established that the affected weld was i
solution heat treated and, therefore, not subject to the IGSCC.
The licensee balieves the cracking may be due to fatigue from flow ' induced vibration.
At this ima the licensee is trying to resolve the problem.
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Paga 2 of 6 At Monticello, IGSCC was c:nfirmed in ona end-ctp-to pipe weld of the 22-inch-diameter distribution header (manifold) and at five welds in the jet pump inlet piping safe ends which are 12 inches in diameter and are made of schedule 80 stainless steel.
The cracks initiated on the inside surface in heat affected zones (HAZs) of the welds.
Some cracks were oriented axially and some circum-farentially.
They varied from is inch to 1 inch in length.
Some axial cracks in the recirculation inlet risers were found to be through-wall during subsequent repair activities and hydrotesting, although ultrasonic examination previously p;rformed on these welds did not reflect this condition.
l At Hatch Unit 1, multiple linear indications characteristic of the IGSCC found at Monticello were identified at seven welds in the large-diameter recirculation and associated residual heat removal (RHR) piping.
The affected welds were t
lccated as follows:
All four 22-inch-diameter manifold end-caps, one 22-inch-diameter branch connection (sweapolet-to-manifold) of the recirculation piping, one elbow-to pipe weld in the 20-inch RHR piping, and one pipe-to pice mild in the 24-inch diameter RHR piping.
The location and orientation of the indications were very similar to these found at Monticello.
The length of the indications ranged un to 33 inch in the axial direction and 13rinch in the 1
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circumfenantial direction.
Based on UT measuresents, the depth of axial comconent of the craci: idicatierts wera found to have essentially penetrated through the wall in three of'th four end-cap welds repairnd to date.
Tha discovery of extensive ICSCC in the large-diameter recirculation piping at Nina Mile Point Unit 1 (NMP 1) after a decade of acceptable service has resulted in increased concern aoout the effectiveness of UT methodology used'in the i
insorsice inspection of stainiess steel SWR pipe weids, particularly in large-diameter piping.
Thersfore, the goal of Item 1 of IES S2-03, Revision 1 was to obtain reassurance of the capability of UT inspection systems, techniques, and operstors to detect significant IGSCC problems in the nine BWR plants that ware performing ISI durin2 fall / winter outages.
The performarca test protocol as stated in Item 1 of IEB 82-03, Revision 1 required the licantee and/or ISI eq:ncies to demo 1 strate their capacility to detect IGSCC in large-diameter recirculat1on system piping be# ore resuming power operation. Within this context,' ~
Electric Power Research Institute's NOE (EPRI-NDE) Center arranged to have five reasonably characterized, service-induced cracked pipe samples from the NMP 1 l
plant available at Battelle Columbus Laboratories (BCL) for industry performance -
l captbility demonstrations (PCDs).
l l
All nine plants have now satisfied the demonstration phase of IEB 82-03, R vision 1.
By letter dated January 28, 1983, EPRI provided each licensee a summary of all teams performances, based on composite results from the five samples, plus a key to identify their ISI team's achievement.
Tha PCD results at BCL have shown that excellent performance can be achieved by l
well trained and experienced personnel with appropriate procedures and evaluation methods.
However, personnel from a relatively few licensee /ISI organizations achieved this level of competence during the first qualification attempt.
The overall results revealed a high failure rate which required retesting of the licensee /ISI organization teams.
Several interrelated factors contributed to
'his rate of failure:
March 4, 1983 b
Page 3 of 8' 1.
UT procedures essentially meeting only the minimum requirements of tHe ASME Section XI code were ineffective.
2.
UT procedures lacked specific. detailed guidance on UT systems and methods proven capable of detecting IGSCC in thick walled piping.
3.
Some UT operators were inexperienced in evaluating signal patterns of reflectors in thick-walled, large-diameter piping.
Thus, some cracks were missed, or were called geometry effects; some geometry effects were falsely called cracks.
4.
Many UT operators, inexperienced about the nature of IGSCC in la.rge-diameter piping, did not establish finite metal path calculations during scanning; this resulted in falsely identified conditions.
In view of the collective results at BCL, a continuation of the PCD program appears necessary.
Accordingly, the EPRI-NDE Center has arranged to have a scries of service-induced cracked specimens available for this purpose at their facility about March 14, 1983.
Tha NRC recognizes that the prescribed actions of this bulletin exceed present plent ISI surveillance requi p ments uncer ASME Code Section XI rules.
- iiowever, in view of the apcarently generic pipe cracking experience' sod results of the i
UT demonstration trials, the NRC believes such an autawntad ISI plan is necassary to reasenaoly assure the integrity of the recirculation system for continued i
4 opsratiors.
These actions are intended to apply only to the currently schedulec rofueling outage for thosa plants listed _in Table 1.
Any licensee who finds these actions will significantly impact the duration of the refueling outage may request ro11ef by writtan request to the appropriate NRC regional office.
Such requests must address (1) the impact.on tae length of the outage, (2) proposed alternative actions, and (3) technical basis for continui'g operation.
n Actions to Be Taken by Licensees of BWR iacilities Identified in Table 1.:.
1.
Befcre resuming power operations following this scheduled or extended outage, the licensee is requested to demonstrate the effectiveness of the.
detecticn capability of the UT methodology planned to be used to examine welds in recircirculation system piping.
It is intended that the demonstrations be performed at the EPRI-NDE Center on service-induced cracked pipe samples made available for this purpose.
Each licensee should assure that the demonstration is valid for the weldments of the recirculation system piping of their plant.
Arrangements should be made to facilitate NRC witnessing of these tests.
The demonstration tests will employ the following criteria.
a.
Ultrasonic Testina System:
To ensure that the field UT system will respond in the same way as the demonstrated system, the same procedures, standards, make and model of the UT instrument, and transducers to be utilized in the plant ISI are to be used in the IGSCC detection capability demonstration.
1
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~~ IEB'83-02 ~
M2roh 4, 1983 J.
V Pag 2 4 of 6 b.
Personnel Performina Demonstration:
UT personnel teams drawn from the licensee /ISI contractor who will be actually supervising, performing examinations, recording data, and evaluating indications at the plant site will participate in the performance demonstration tests.
All members of the teams must participate directly in the UT scanning, data recording, and evaluation of the test samples.
To ensure completion of testing within the time constraints below, the team should be limited to six persons.
For subsequent plant inspections, the personnel / equipment requirements noted below will apply.
c.
Pipe Samples:
The total number of pipe samples selected should constitute an equivalent of 120 inches of weld for the demonstration tests.
d.
Acceptable criteria:
Eighty percent of the total number cf preselected cracks in the sample control group must be called correctly to constitute an acceptable test.
Excessive false call rates may result in an unacceptable performance rating.
a.
Demonstration Time Limit:
ALARA radiation dosa considerations place constraints upon tie time spent in field inservice inspection of a weld.
Therefore, a time limit of six hours, not including equipment calibration time, will be imposed for the examination and data recording.
Completion of data evaluation and praparation of final results of individual licenses /ISI contractors should take no longer than one additional working day.
f.
Review of UT Procedures:
The specific proceoure(s) to be used by the licensee /ISI contractor (s) for plant inservice inspecticn. is to be f
made available for review as part of the demonstration a~ctivity.
It is expected that the UT procedure and equipment system will have been l'
validated to bs capable of detecting IGSCC by the licensee /ISI con-tractor before initiating the scheduled demonstration activities.
NOTE:
Some of the licensees listed in Table 1 have completed efforts to validate the UT detection capability to be used to perform plant inspections in accordance with ths requirements of Action Item I of IEB 82-03, Revision 1.
These licensees need not repeat this effort in accordance with Action Item 1 of this bulletin provided that:
the previous validated inspection group performs the new plant examination using identical UT procedures, standards, make and model of UT instrument, and the same make and model transducers that were used to complete the previous validation effort. In addition, the UT personnel employed in the new examination must be the same; or those having l
appropriate training (documented) in IGSCC inspection using cracked thick-wall pipe specimens, and are under direct supervision of the Level II/III UT operators who successfully complete the performance demonstration tests.
IES 83-0.2
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7' March 4,1983 Page 5 of 6 2.
Before resuming power operations licensees are to augment their ISI programs to include an ultrasonic examination of the following minimum number of recirculation system welds:*
Ten welds in recirculation piping of 20-inch diameter, or larger.
a.
b.
Ten welds of the jet pumps inlet riser piping and associated safe-ends.
c.
Two sweepolet-to-header (manifold) welds of jet pump risers nearest the end caps, if applicable to the design.
If flaws indicative of cracking are found in the above examination, additional inspection is to be conducted in accordance with IWB 2430 of ASME Code Section XI.
3.
Before resuming power operations following the outage, the licensee is to report the results of the Item 2 inspection and any corrective actions (in the event cracking is identified).
This repnrt should also include the susceutibility matrix used as a baris for welds selectea for e.samination (e.g. stress rule index, carton content, high stress location, repair history) anc their valugs for each weld examined.
4.
The NRC has an on going program to evaluata possibis additional longer-tartt requirements relatt is to the IGSCC pruoles in the BWR recirculation system piping.
The NRC may neeo additional infern.ation as part of this program.
Therefore, licandees are requested to retair, the records and data developed pursuant to the inspections performed in accordance with Item 2.
5.
The written report required by Itsa 3 shall be submitted to the aparepriate r
Pegional Aaministrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1354, as amended.
The original copy of tne cover lettar and a cosy of the reports shall ba transmitted to the U.S.
Nuclear Regulatory Commission, Document Control Desk, W<ashington, D.C. 20555 for reproduction and distribution.
This request for information was approved by the Office of Management and Budget under clearance number 3150-0096 which expires 12/31/84.
Comments on burden and duplication should be directed to the Office of Management and Budget, Rsports Management, Room 3208, New Executive Office Building, Washington, 0.'C.
20503.
"Since Big Rock Point and Lacrosse do not have jet pumps, the licensees of these plants should provide an equivalent sanpling of the recirculation pioing system based on the plant design.
/
IEB 83-02 March 4; 1983 Paga 6 of 6 Although no specific request or requirement is intended, the following information would help the NRC evaluate the cost of implementing this bulletin:
Staff time to perform requested demonstration.
Staff time to prepare written responses.
The occupational radiation exposure experienced.
If you have any questions regarding this mattar, please contact the Regional Administrator of the appropriata NRC Regional Office dr one of the technical contacts listed below.
Richard C. DeYoung, D.irector Office of !nspection and Enforcement Technic.tl
Contact:
William J. Collins, IS 492-7275, Warren Hozelton, NPR 492-8075 Attachaents:
1.
Table 1 2.
Table 2
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List of Recently~ Issued IE Bulletins 9
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~ Attachment I t-IEB 83?02 March 4, 1983 Table 1 BWR Plants Scheduled to be in the Next Refueling Mode or Extended Outage After January 31, 1983 LICENSEE PLANT RELOAD DATE Philadelphia Electric Co.
Peach Botton Unit 3 February 1983 Vcrmont Yankee Nuclear Power Vermont Yankee March 1983 Company Tenessee Valley Authority Browns Ferr/ Unit 1 March 1583 N:traska Public Powr District Ccoper April 1993 Gzorgia Pewer Co.
Hatch Unit 2 April 1983 Consumers Powr Co.
Big Rock Point May 1983 Power Authority of the State FitzPatrick May 1983 of New York Commonwsalth Edison Co.
Quad Cities Unit 2 August 1983 l
Tcnr, esse.e Valley Authority Browns Ferry linit 3 September 1983 Carolina Power a Light Co.
Brunswick Unit 2 September 1983 Dairyland Power Corp.
Lacrossa October 1983 i
Philadelphia Electric Co.
Peach Botton Unit 2 October 1983 Commonwealth Ed'ison Co.
Dresden Unit 3 October 1983 Boston Edison Co.
Pilgrim Unit 1 January 1984
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A.ttichment 2 IEB 83-02 March 4, 1983
- TABLE 2 CRACK INDICATIONS IN BWR RECIRCULATED SYSTEM PIPING PLANT PIPE SIZE WELD LOCATION HOW DETECTED NMP 1*
28" Dia.
Pipe to safe ends Initial crack-visual Visual leakage 28" Dia.
Pipe to Pipe UT 28" Dia.
Pipe to pump casing Visual - UT Monticello 12" Dia.
Riser to Safe End Leakage (weepage) 12" Dia.
Riser to Safe End Weepage - UT 12" Dia.
Riser to Safe End Weepage - UT 12" Dia.
Riser Elbow to Pipe UT 22" Dia.
Elbow to. Pipe Leakage (weepage)
During Hydrotest-visua!
Hmteli 20" Dia.
Elbow to Pipe (RHR)
UT 22" Dia.
Manifold End cap tfr 22" Dia.
Pipe to P1pe (RHR)
UT 22" Dia.
12" Hisar Sweepclet to-Manifold UT 1
Browns Ferry 22" Dia.
12" Riser Sweapolet to Manifold UT 22" Dia.
12" Riser Sweapolet to Manifold UT 1
Brunswick 28" Dia.
Elbow to Pipe UT 12" Dia.
Riser to Safe End Leakage (weeraga) 12" Dia.
Riser to Safe End Leakage (Weepage)
Dresden 2 28" Dia.
12" Dia.
Riser Pipe to Elbow UT Footnotes:
Cracks were found in 90% of welds examined Generally, there were indications of more than one axial or circumferential aligned crack in each affected weld.
IEB 83-02 March 4, 1983 LIST OF RECENTLY ISSUED IE BULLETINS Bulletin Date of No.
Subject Issue Issued to 83-01 Failure of Reactor Trip 02/25/83 All PWR facilities Breakers (Westinghouse 08-50) holding an OL and to Open on Automatic Trip other power reactor Signal facilities for information 82-04 Deficiencies in Primary Con-12/03/82 All power reactor tainment Electrical Pene-facilities holding tration Assemblies an OL or CP 82-03 Stress Corrosion Cracking in 10/28/82 Operating BWRs in
>!av, 1 Thick-Wall Large-Diameter Table 1 for action Stainless *,tael, Facircula-and other OLs and cps tion Systen Ploing at SWR for information 01 ants S2-03 Stress Corrosion' Cracking in 10/14/82 Operating BWRs in Thick-Wall Large-Diametsr, Table 1 for action Stainless Steel, Recircula-and other OLs and cps tion System Piping at BWR for information Plants 82-01 Alteration of Radiographs of 03/18/82 All power reactor Rav 1, Weids in Piping Subassemblies facilities with Supp 1 an OL or CP l
82-02 Degradation of Threaded 06/02/82 All PWR facilities Fastenera in the Reactor Coolant with an OL for
+
Pressure Boundary of PWR olants action and all other OLs or cps for information 82-01 Alteration of Radiographs of 05/07/82 All power reactor RGv. 1 Welds in Piping Subassemblies facilities with an OL or CP 82-01 Alteration of Radiographs of 03/31/82 The Table 1 Welds in Piping Subassemblies facilities for action and to all others. for information 81-02 Failure of Gate Type Valves 08/18/81 All power reactor Supplement to Close against Differential facilities with an 1
Pressure OL or CP OL = Operating License CP = Construction Permit l
ATTACHiENT 3 0
TRANSCRIPT OF 'THE"CONMISSION'S
" MEETING WITH INDUSTRY -- BWR OWNER'S GROUP,"
(Chairman Palladino)
JULY 15,1983 101 1
them to those dates, but on closer examination seeing what' they can do to meet those da:es.
3 well, let's see if we can articulate your 4
proposal. You proposal is to wait for the data that is due 5
August 4th and then review the situation.
In the interim 6
it suggests the staff sit down*ita each of the licensees 7
and see what ks.nd of a realistic schedule can be developed a
for the inspection with the idea of accelerating the 9
inspection as muen as pcssible.
10 Is that a fair representation of your 11 proposal?
12 COMM.ISSIGNER G3LINSKY:
Well, yes, and he has 13 ample authority on the. basis of tant August 4th resul: to l
14 act in the way he proposed right now.
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15 MR. DENTON:
And there will be more field l
M a::;erience by August 4tn.
- think what we have been doing
- C.
17 is watching this c11 in and the nu.noer of PN's stack up I
18 and up and are trying to decide when we snould move 19 faster.
20 CHAIRMAN PALLADINO:
Cf course, we are no:
21 taking away from liarold centon any authority ne has := ac:
l i
=
in. emergencies as he sees fit.
3 COMMISSIONER GILINS15Y: August 4:n would in any l
24 case be before the date that any of these inspections 3
would taxe place even if they were ordered rign now.
TAYLCE ASSOCIATES 1625 i STREET. N.W. - SUITI 1004 WASHINGTON. C.C.
20006 I
(202) 293 3950
_