ML17053B180
| ML17053B180 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/29/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Rhode G NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 7911150051 | |
| Download: ML17053B180 (24) | |
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Docket No. 50-410 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA19406 OC7 2 9
)97'iagara Mohawk Power Corporation ATTN:'r.
G.
K.
Rhode Vice President System Project Management 300 Erie Boulevard, West
- Syracuse, New York l3202 Gentlemen:
The enclosed IE Bulletin 79-17, Revision 1, is forwarded to you for information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, Director
Enclosures:
1.
IE Bulletin No. 79-17, Revision 1 w/Attachment 2.
List of IE Bulletins Issued in the Last Six Months CONTACT:
L.
E. Tripp 215-337-5282 cc w/encls:
Eugene B. Thomas, Jr.,
Esquire i'OXXX50O g ~
ENCLOSURE 1 UNITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCEMENT 7908220&7 i 3 MASHINGTON, D. C.
20555 IE Bulletin No. 79-17 Revision 1
Date:
October 29, 1979 Page 1 of 5 PIPE CRACKS IN STAGNANT BORATED MATER SYSTEMS AT PWR PLANTS Description of Circumstances:
IE Bulletin No. 79-17, issued July 26, 1979, provided information on the cracking experienced to date in safety-related stainless steel piping systems at PWR plants.
Certain actions were required of all PMR facilities with an operating license within a specified 90-day time frame.
Rl Rl Rl Rl After several discussions with licensee owner group representatives and Rl inspection agencies it has been determined that the requirements of Item 2, Rl particularly the ultrasonic examination, may be impractical because of un-Rl availability of qualified personnel in certain cases to complete the in-Rl spections within the time specified by the Bulletin. To alleviate this Rl situation and allow licensees the resources of improved ultrasonic inspec-Rl tion capabilities, a time extension and clarifications to the bulletin have Rl been made.
These are referenced to the affected items of the original Rl bul leti n.
During the period of November 1974 to February 1977 a number of cracking incidents have been experienced in safety-related staihless steel piping systems and por-tions of systems which contain oxygenated, stagnant or essentially stagnant bor-ated water.
Metallurgical investigations revealed these cracks occurred in the weld heat affected zone of 8-inch to 10-inch type 304 material (schedule 10 and 40), initiating on the piping I.D. surface and propagating in either an inter-granular or transgranular mode typical of Stress Corrosion Cracking.
Analysis indicated the probable corrodents to be chloride and oxygen contamination in the affected systems.
Plants affected up to this time were Arkansas Nuclear Unit 1, R.
E. Ginna, H.
B. Robinson Unit 2, Crystal River Unit 3, San Onofre Unit 1, and Surry Units 1 and 2.
The NRC issued Circular No. 76-06 (copy attached) in view of the apparent generic nature of the problem.
During the refueling outage of Three Mile Island Unit 1 which began in February of this year, visual inspections disclosed five (5) through-wall cracks at welds in the spent fuel cooling system piping and one (1) at a weld in the decay heat removal system.
These cracks were found as a result of local boric acid buildup and later confirmed by liquid penetrant tests.
This initial identification of cracking was reported to the NRC in a Licensee Event Report (LER) dated May 16, 1979.
A preliminary metallurgical analysis was performed by the licensee on a section of cracked and leaking weld joint from the spent fuel cooling system.
Rl - Identsfles those additions or revision to IE Bulletin No. 79-17
Enclosure 1
IE Bulletin No. 79-'7 Revision 1
Date:
October 29, 1979 Page 2 of 5 The conclusion of this analysis was that cracking was due to Intergranular Stress Corrosion Cracking (IGSCC) originating on the pipe I.D.
The cracking was localized to the heat affected zone where the type 304 stainless steel is sensitized (precipitated carbides) during welding.
In addition to the main through-wall crack, incipient cracks were observed at several locations in the weld heat affected zone including the weld root fusion area where a miniscule lack of fusion had occurred.
The stresses responsible for cracking are believed to be primarily residual welding stresses in as much as the calculated applied stresses were found to be less than code design limits.
There is no conclusive evidence at this time to identify those aggressive chemical species which promoted this IGSCC attack.
Further analytical efforts in this area and on other system welds are being pursued.
Based on the above analysis and visual leaks, the licensee initiated a broad based ultrasonic examination of potentially affected systems utilizing special techniques.
The systems examined included the spent fuel, decay heat removal, makeup and purification, and reactor building spray systems which contain stagnant or intermittently stagnant, oxygenated boric acid environments.
These systems range from 2 1/2-inch (HPCI) to 24-inch (borated water storage tank suction),
are type 304 stainless
- steel, schedule 160 to schedule 40 thickness respectively.
Results of these examinations were reported to the NRC on June 30, 1979 as an update to the May 16, 1979 LER.
The ultrasonic inspection as of July 10, 1979 has identified 206 welds out of 946 inspected having UT indications characteristic of cracking randomly distributed throughout the aforementioned sizes (24"-14"-12"-10"-8"-2" etc.) of the above systems.
It is important to note that six of the crack indications were reportedly found in 2 1/2-inch diameter Rl pipe of the high pressure injection lines inside containment.
These lines are attached to the main coolant pipe and are nonisolable from the main coolant system except for check valves.
All of the six crack indications were found in two Rl high pressure injection lines containing stagnated borated water.
No crack Rl indications were found in high pressure injection lines which were utilized for Rl makeup operations.
Recent data reported from Three Mile Island Unit 1 indicates that the extent Rl of IGSCC experienced in stainless steel piping, at that facility may be more Rl limited than originally stated above.
Of the 1902 total welds originally Rl inspected 350 contained U.T. indications which required further evaluation.
Rl These 350 welds have been,reinspected with a second U.T. procedure which pur-Rl portedly provides better discrimination between actual cracks and geometrical Rl reflectors.
- Hence, the licensee now estimates that approximately 38 of the Rl 350 welds contain IGSCC and the remaining welds, including those in high pressure Rl injection and decay heat lines, contain only geometrical reflectors.
Further Rl metallurgical analysis of these welds is required to verify the adequacy of the Rl U.T. procedures and to determine the nature of the cracking.
Rl
Enclosure 1
IE Bulletin No. 79-17 Revision 1
Date:
October 29, 1979 Page 3 of 5 For All Pressurized Mater Reactor Facilities with an Operating License:
1.
Conduct a review of safety related stainless steel piping systems within 30 days of the date of this Bulletin (July 26, 1979) to identify systems and portions of systems which contain stagnant oxygenated borated water.
These systems typically include ECCS, decay/residual heat removal, spent fuel pool cooling, containment spray and borated water storage tank (BWST-RWST) piping.
Rl For this revie~, the term "stagnant, oxygenated borated water systems" refers Rl to those systems serving as engineered safeguards having no normal operating Rl functions and contain essentially air saturated borated water where dynamic Rl flow conditions do not exist on a continuous basis.
However, these systems Rl must be maintained ready for actuation during normal power operations.
Mhere Rl your definition for stagnant differed from the one given above please supple-Rl ment your previous response within 30 days of this Bulletin revision.
Rl (a)
Provide the extent and dates of the hydrotests, visual and volumetric examinations performed per 10 CFR 50.55a(g)
(Re:
IE Circular No. 76-06 attached) of identified systems.
Include a description of the non-destructive examination procedures, procedure qualifications and accep-tance criteria, the sampling plan, results of the examinations and any related corrective actions taken.
(b)
Provide a description of water chemistry controls, summary of chemistry
- data, any design changes and/or actions taken, such as periodic flushing or recirculation procedures to maintain required water chemistry with respect to pH, B, Cl-, F-, 02.
(c)
Describe the preservice NDE performed on the weld joints of identified systems.
The description is to include the applicable ASME Code sec-tions and supplements (addenda) that were followed, and the acceptance criterion.
(d)
Facilities having previously experienced cracking in identified systems, Item 1, are requested to identify (list) the new materials utilized in repair or replacement on a system-by-system basis.
If a report of this information and that requested above has been previously submitted to the NRC, please reference the specific report(s) in response to this Bulletin.
2.
All operating PWR facilities shall complete the following inspection on the Rl stagnant piping systems identified in Item 1 at the earliest practical date Rl not later than twelve months from the date of this bulletin revision.
Fa-Rl cilities which have been inspected in accordance with the original Bulletin, Rl Sections 2(a) and 2(b) satisfy the requirements of this Revision.
Rl
Enclosure 1
IE Bulletin No. 79-17 Revision 1
Date:
October 29, 1979 Page 4 of 5 (a)
Until the examination required by 2(b) is completed a visual examination shall be made of all normally accessible welds of the engineer'ed safety systems at least monthly to verify continued systems integrity.
Sim-ilarly, the normally inaccessible welds, shall be visually examined during each cold shutdown.
The relevant provisions of Article IMA 2000 of ASME Code Section XI and Article 9 of Section V are considered appropriate and an acceptable basis for this examination.
For insulated piping, the examination may be conducted without the removal of insulation.
During the examination particular attention shall be given to both insulated and noninsulated piping for evidence of leakage and/or boric acid residues which may have accumulated during the service period preceding the examination.
Mhere evidence of leakage and/or boric acid residues are detected at, locations, other than those normally expected, (such as valve stems, pump seals, etc.) the piping shall be cleaned (including insulation removal) to the extent necessary to permit further evaluation of the piping condition.
In cases where piping conditions observed are not sufficiently definitive, additional inspections (i.e., surface and/or volumetric) shall be conducted in accordance with Item 2.(b).
(b)
An ultrasonic examination shall be performed ona representative sample of circumferential welds in normally accessible portions of systems identified by 1 above.
It is intended that the sample number of welds selected, for examination include all pipe diameters within the 2 1/2-inch to 24-inch range with no less than a 10 percent sampling being taken.
The approach to selection of the sample shall be based on the following criteria:
(1)
Pipe Material Chemistry - As a first consideration, those welds in austenitic stainless steel piping (Types 304 and 316 ss) having 0.05 to 0.08 wt. X carbon content based on available material certification reports.
(2)
Pipe Size and Thickness - An unbiased mixture of pipe diameters and actual wall thickness distributed among both horizontal and
,vertical piping runs shall be included in the sample.
(3)
System Importance - The sample welds shall focus the examination primarily on those systems required to function in the emergency core cooling mode and secondly, on the containment spray system.
The U.T. examination sample may be focused on noninsulated piping runs.
The evaluation shall cover the weld root fusion zone and a
minimum of 1/2 inch on the pipe I.D. (counterbore area) on each side of the weld.
The procedure(s) for this examination shall be essentially Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Normally accessible refers to those areas of the plant which can be entered during reactor operation.
Rl
'Rl
Enclosure 1
IE Bulletin No. 79-17 Revision l Date:
October 29, 1979 Page 5 of 5 in accordance with ASME Code Section XI, Appendix III and Supplements Rl of the 1975 Winter Addenda, except all signal responses shall be eval-,
Rl uated as to the nature of the reflectors.
Other alternative examination Rl
- methods, combination of methods, or newly developed techniques may be Rl used provided the procedure(s) have a proven capability of detecting Rl stress corrosion cracking in austenitic stainless steel piping.
Rl For welds of systems included in the sample having pipe wall thickness Rl of 0.250 inches and below, visual and liquid penetrant surface examina-Rl tion may be used in lieu of ultrasonic examination.
Rl (c) If cracking is identified during. Item 2(a) and 2(b) examinations, all Rl welds in the affected system, shall be subject to examination and repair Rl considerations.
In addition, the sample welds to be examined on the Rl remaining normally accessible noninsulated piping shall be increased to Rl 25 percent using the criteria outlined in paragraph 2(b).
In the event Rl that cracking is identified in other systems at this sampling level, Rl all accessible and inaccessible welds of the systems identified in Rl item 1 shall be subject to examination.
Rl 3.
Identification of cracking in one unit of a multi-unit facility which causes safety-related systems to be inoperable shall require immediate examination of accessible portions of other similar units which have not been inspected under the ISI provisions of 10 CFR 50.55a(g) unless =justification for con-tinued operation is provided.
4.
Any cracking identified shall be reported to the Director of the apppropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification followed by a 14 day written report.
Provide a written report to the Director of the appropriate NRC Regional Office within 30 days of the date of this bulletin revision addressing the results of your review if required by Item 1.
Provide a schedule of your inspection plans in response to Item 2(b) in those cases in which the inspections have not been completed.
R1 R1 Rl Rl Rl 6.
Provide a written report to the Director of the appropriate NRC Regional Rl Office within 30 days of the date of completion of the examinations required Rl by Items 2(a), 2(b), or 2(c) describing the inspection results and any cor-Rl rective actions taken.
Rl 7.
Copies of the reports required by Items above shall also be provided to the Director, Division of Operating Reactors, Office of Inspection and Enforce-ment, Washington, D. C.
20555.
Approved by GAO, B180225 (R0072), clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
Attachment:
IE Circular No. 76-06
Attachment to -IE Bulletin No. 79-17 Revision 1, Havmber 26, 1976 M Circular Ha-76-06 STRKBS CORROSION CR'CES EH STAGNAN~, LOP PRMSURE STAXÃLZSS PIPING COHTAXNXhC MRXC ACID SOLUTEWil AT PWR's DESCRIPTIONS'F C RCC.'KXAHCM~:
During the period Havcmber 7, 1974 to November 1, 1975, seve-,.al incidents of through-wall cra"keg have occurred in the 10-inch, schedvQ,e IQ type 304 stainless steal piping of the Reacto-Bui'tding Spray and Decay Peat Rettova3 Systems at Arhnsas Huclea" Plant Ho. l.
On October 7, 1976, US.rginia Electric and Pat"er also reported thraush-
'vM~
cracking Xn the 10>>inch schedule
<0 type 304 stateless dischazS piping of the "A" recixculat" on spray heat exchanger st Suzzy. U"'t Ho. 2.
A recent inspection af Unit 1 Contain~ent Recirculation Spray Piping revealed craig s&Q.az to Unit 2.
On October 8, l976, another incident of similar cracking in 8-inch schedule 10 type 304 stainless piping of the S"fcty 'Ln)ection Pu"p Suction Line.<<t the Cf.nna facility vas repaztad by the licensee.
Znfaination received, an th metallurgical analysis conducted ta d te indicates that toe failures vera the result of inte granular tz ss corros an crac~-
that init'atcd on the inside of the piping.
A co~="nag.ity of factors observed associated 'ith the corrosion
=.
cha"..isa wez'Q ~
The cracks vere adjacent ta and propagated along t~eld zones af the thin-t=alled 3.at'ressure pipi.g, not part af the reactor coa'a"t.
system.
2.
Cracking occu ed in pipittg containing re1ativcly s agnant bor'c acid so'ntian rat required Ear non@I operatic,g conditions.
Analysis af surface products at thf.s time indicate " chloride ion intezartian vith o+ide farmUon in the relatively stagnant.boric acid alu"ior. as the prob blc carrod nt, t:ith the state, af stress probably due to t"elding and/ar fabricatian The source of "he chloride ion is ttot definitely kno~~.
Hot~ver, at ABO-1 the chic~ca and sulfide 1.eve1 observed in the su faca ta..ish filrtnear welds is beLieved to have bean introduced into the piping during tasting af the sodium tiiosul ate discharge valves, o" valve leakage.
S~a ly, e-Gin"a the chlorides and potential az~gcn
I XE Q,rculaz Ho. 76-06 November 26, l976 availabilit7 were assumed to have been pzesent since original construction af the bozatcd water storage tank which is vented ta atmosphere.
Corrosion attack at Surry is attributed to in-leahage of chlorides through recircu>ation "prey heat exchange tubing. alla~aug buildup of contauduated water i-an atherWsc normally dry spray piping.
AGTXOH TO 3E TAZr> HY LICMSKZ:
l; Provide
- a. description a" your program fa" assuring continued
~ itLtegzi~ af those safe~-related piping systems w'nich ere no>>
frequently flushed, ar which cont"in nonfla~~ng, liquids.
This pzagzam should include consideration of hydrostatic testing in
~ccazd~ce Wth ASM Code Section XI rules (1974 Fditian) i'az aU. active systems zequX.red far safety injection and containment
- spray, includi=g their recirculation made, from source of water supply up to the ecaad isolation valve af the primary s>atem.
.Solar tests should be considered foz other safety-relat d pipin Spat~~
~
2.
Your program shauld also consider volumetric examination of s.
representative nM>>er af'circumferential pipe welds by nan-8estzuctivo e'aminatien techniques.
Such exam nations should be: performed generally in acco".dana~ with Appendix I af Sec Son XX of the ARK Cade, except "hat the examined area shoula cover a distance of approximately six. (6) times the p" pe wa13. thickness (bu-not less than 2 inches a..d need not exceed 8 inches) on each side of t2Ie weld.
Supplenenta~
ex~ation techniques, such
&s radiagrap2iy, should be used:
Mhere necessary for evaluation or confiz=ation af u3.trasanic indications resulting fzon such exam&ation.
3.
A report describing your program an" schedule far these inspec-tions should be submitted within 30 days after receip" of this CS,zMwax' 4.
The, 2lRC Regional Office should be infazmed within Zli hours, af any adverse findings resulting du"ing nondestructive evaluation of the accessible piping zelda identified above.
5, A sumnaxp rcport of the examinations and evaluation of resets should be submitted within 60 day from the date oZ completion of pzaposed testing and examinations.
XE Circular Ho. 76W6 November 26, l976 TkLs scraaary report should also include a brief description of plant conditions, operating procedures or other activities which provide assurance that the r/fluent chemist~ vfLX maintain los levels af potential corrodants in such relatively stagnant regions Mthin the piping.
Yau responses should be submitted to the Director of tMs office,
~4th a copy to the NRC Office of Inspection and Enforcement, Dh+sian of Reactor Inspectian
- Programs,
'hashington, D.C.
20555.
Approval af NRC requirements for reports coacting passible generic problems has been obtained under 44 U. S.C 3152 f"om the Q.S..
Qe ez 1
Accounting Office.
(CAQ Appro'val 8-180255 (R0062), n;pires 7/31(77.)
Bulletin Subject No.
ENCLOSURE 2 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Date Issued IE Bulletin No. 79-17 Revision.No.
1 Date:
October 29, 1979 Page 1 of 3 Issued To 79-10 Requalification Training 5/11/79 Program Statistics All Power Reactor Facilities with an OL 79-11 79-12 Faulty Overcurrent Trip Device in Circuit Breakers for Engineered Safety Systems Short Period Scrams at BWR Facilities 5/22/79 5/31/79 All Power Reactor Facilities with an OL or CP All GE BWR Facilities with an OL 79"01A Environmental qualification 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental qualification of ASCO Solenoid Valves)
All Power Reactor Facilities with an OL or CP 79-02 (Rev 1) 79-13 79" 14 Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts Cracking in Feedwater System Piping Seismic Analysis for
=As-Built Safety Related Piping Systems 6/21/79 6/25/79 7/2/79 All Power Reactor Facilities with an OL or CP All PWRs with an OL (for Action),
All Other Power Reactor Facilities with an OL or CP (For Information)
All Power Reactor Facilities with an OL or CP
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
IE Bulletin No. 79-17 Revision No.
1 Date:
October 29, 1979 Page 2 of 3 Bulletin No.
Subject Date Issued Issued To 79-15 Deep Draft Pump Defi-7/ll/79 ciencies All Power Reactor Facilities with an OL or CP 79-14 (Revision 1)
Same Title as 79-14 7/18/79 Same as 79-14 79-16 79-17 Vital Area Access Con-7/30/79 trois Pipe Cracks in Stagnant 7/26/79 Borated Water Systems at PWR Plants All Holders of and Applicants for Reactor Operating Licenses All PWR Power Reactor Facilities with an OL 79-05CK06C Nuclear Incident at Three Mile Island-Supplement 7/26/79 All PWR Power Reactor Facilities with an OL 79-18 Audibility Problems 8/7/79 Encountered on Evacuation All Power Reactor Facilities with an OL 79-19 79-20 Packaging Low-Level Radioactive Waste for Transport and Burial Same Title as 79-19 8/10/79 8/13/79 All Power and Re-search Reactors with OL, all Fuel Facilities (except Uranium Mills),
and certain Materials Licensees Cer tain Materials Licensees 79-21 Temperature Effects on 8/13/79 Level Measurements All Power Reactor Facilities with an OL or CP
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
IE Bulletin No. 79-17 Revision No.
1 Date:
October 29, 1979 Page 3 of 3 Bulletin No.
Subject Date Issued Issued To 79"02 (Rev 1)
(Supplement No.
1)
Same Title as 79-02 79-14 Same Title as 79-14 (Supplement) 8/15/79 8/20/79 Same as 79-14 Same as 79-02 (Rev 1) 79" 13 (Rev 1)
Cracking in Feedwater System Piping 8/30/79 All Designated Applicants for OLs 79-22 Possible Leakage of Tubes 9/5/79 of Tritium Gas Used in Timepieces for Luminosity Each Licensee who Receives Tubes of Tritium Gas in Timepieces for Luminosity 79"14 (Supplement No. 2) 79-23 79-24 Same as Title 79-14 Potential Failure of Emergency Diesel Generator Field Exciter Transformer Frozen Lines 9/7/79 9/12/79 9/27/79 Same as 79-14 All Power Reactor Facilities with an OL or CP All Power Reactor Facilities which have either OLs of CPs and are in late stage of construction 79-13 (Rev.
2)
Cracking in Feedwater System 10/17/79 Piping All PWRs with an OL and Designated Applicants (for Action), All Other Power Reactor Facilities with an OL or CP (for Information)
/
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