ML17272B060
| ML17272B060 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/29/1979 |
| From: | Beverly Smith NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | |
| Shared Package | |
| ML17272B061 | List: |
| References | |
| NUDOCS 7908220157 | |
| Download: ML17272B060 (13) | |
Text
UNITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accession, No.:
OFFICE OF INSPECTION AND ENFORCEMENT 7908220157 WASHINGTON, D.C.
'20555 October 29, 1979 IE Bulletin Ho. 79-17 Revision 1
PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS Descri ption of Ci rcums tances:
IE Bulletin No. 79-17, issued July 26,
- 1979, provided information on the. cracking Rl experienced to date in safety-related stainless steel piping systems at PWR Rl plants.
Cer tain actions were required of all PWR facilities with an operating R1 license. within a specified 90-day time frame.
Rl After several discussions with licensee owner group representatives and inspection R1 agencies it has been determined that the requirements of Item 2, particularly R1 the ultrasonic examination, may be impractical because of unavailability -of R1 qualified personnel in certain cases to complete the inspections within the time R1 specified by the Bulletin.
To alleviate this situation and allow licensees the Rl resources of improved ultrasonic inspection capabilities, a time extension and Rl clarifications to the bulletin have been made.
These are referenced to the Rl affected items of the original bulletin.
Rl During the period of November 1974 to February 1977 a number of cracking incidents have been experienced in safety-related stainless 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 (veld 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 Onofr e Unit 1, and Surry Units 1 and 2.
The NRC issued Circular Ho. 76-06 (copy enclosed} 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 - Id n:
e '.hosea".'tions or revision o
IC Bulletin No.
79-17
IE Bulletin No.. 79-1/
Revi'sion 1
October 29, 1979 Page 2 of 5
'he 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 sensi tized (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 usion nad 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 ste l, schedule 160 to schedule 40 thickness respectively.
Results of these examinations were reported to the HRC on June 30, 1979 as an update to the Hay 16, 1979 LER.
The ultrasonic i nspection 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 s'tagnated borated water.
No crack R1 indications were found in high pressure injection lines which were utilized for Rl makeup operations.
Recent data reported from Three f/ile Island Unit 1 indicates that the extent Rl of IGSCC experienced in stainless steel piping at that facility may be more R1 limited than originally stated above.
Of the 1902 total welds originally R1 inspected 350 contained U.T. indications. which required further evaluation.
R1 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 R1 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
IE Bulletin Ho.
79-17 Revision 1
October 29, 1979 Page 3 of 5 For All Pressurized Mater Reactor Facilities wi th an Ooerating License:
1.
Conduct a review of safety related stainless steel piping system's within 30 days of the date of this Bulletin (July 26, 1979) to identify systems and portions of systems which con.ain 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-Rl)ST) piping.
For this review, the term "stagnant, oxygenated borated water systems" refers Rl to those systems serving as engineered safeguards naving no normal cperating R1 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 normai power operations.
Where 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 enclosed) 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 ASNE 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 HRC, please reference the specific report(s) in response to this Bulletin.
2.
All operating PHR facilities shall complete the following inspection on the Rl'tagnant piping systems identified in Item 1 at the earliest practical date
- R1 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 o> this Revision.
Rl
IE Bulletin iso.
79-17.
Revision I October 29, 1979 Page 4 of 5 (a)
Until the examination required by 2(b) is completed a visual examination Rl shall be made of all normally accessible welds of the engineered safety Rl syst ms at least monthly to verify continued systems integrity.
Sim-Rl ilarly, the normally inaccessible welds, shall be visually examined Rl during 'each cold shutdown.
Rl (b)
The relevant provisions of Article IHA 2000 of ASi'lE Code Section XI and Article 9 of Section ll are considered appropriate and an acceptable basis for this examination.
For insulated piping, the examination may be conducted without th 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.
Hhere 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).
An ultrasonic examination shall be performed on~a representative sample of circumferential welds in normally accessible
- portions of systems identified by I above.
It is intended that the sample number of welds selected for examination include all pipe diameters within the 2 I/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:
(I)
Pipe i~1aterial Chemistry - As a first consideration, those we1ds in austenitic stainless steel piping (Types 304 and 316 ss) having 0.05 to 0.08 wt.
% 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 I/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 RI RI Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl R'1
~,'!or~-.a:'iy ac essib!e i e=.ers
~o those areas o
the plant whic?
can be nzerea durl;)9 reac i ol opet a~ ~on
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IE Bulletin >/o. 79-17 Revision 1
October 29, 1979 Page 5 of 5 in accordance with AS)/E Code Section XI, Aopendix III and Suoplements R1 of the 1975 Hinter Addenda, except all signal responses shall be eval-R1 uated as to the nature of the reflectors.
Other alternative examination Rl
- methods, combination of methods, or newly developed techniques may be R1 used provided the procedure(s) have a oroven capability of detecting R1 stress corrosion cracking in austenitic stainless steel pioing R1 ll For welds of systems included in the sample havinq pioe 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 durinq Item 2(a) and 2(b) examinations, all R1 welds in the affected
- system, shall be subject to examination and repair Rl considerations.
In addition, the sample welds to be examined on the R1 remaininq normally accessible noninsulated pipinq shall be increased to R1 25 oercent using the criteria outlined in oaragraph 2(b).
In the event R1 that cracking i s 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.
R1 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 ooeration is orovided.
4 5.
6.
Any cra"king identified shall be reported to the Director of the apporooriate
'RC 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 t!RC Regional R1 Office within 30 days of the date of this bulletin revision addressing the Rl results of your review i required by Item 1.
Provide a schedule of your Rl inspection plans in response to Item 2(b) in those cases in which the Rl inspections have not been completed.
R1 Provide a written report to the Director of the appropriate HRC Pegional R1 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-R1 rec.ive actions taken.
R1 7.
Cooies of the resorts required by Items above shall also be provided to the Director, Division of Operating Reactors, Offic'e of Inspection and Enforce-ment, Mashington, D.C.
20555.
Aporoved bv ~'0, 8180225
(.0072), clearance exni>es 7/M/80.
Approval was
".iven u~.'er
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Hoveaber,26, 1976
- ",=- Circular Ho '6-06 STB:SS CORROSXOh CR'CM XH STAGNAh, LOP PRESSi$3:" STXZHXZSS P&L1C CC5TJG2iLNG ROE'ZC ACXD SOLUTZON AT PVR" DZSCRXPTXOH OF CiP.CC.'KTAHCES:
l3urMg tbe period 3ovenber 7, "974 to Navenber 1, 1975, eve-,al inc&ent of through-wal1 cxackQxg 'h~e occur ed In the 10-inch, chedule 10 type 304 stagese steel pip~mg of the Reactor Building Spray and Decay B~t Re-ova3 Syst~ a" Arkansas hucl~"r Plant?lo. l.
On October 7, 1976, VS.rginia Zlectric snd Pover al o repc ted "hraugb-
'well cracking Xn the 3.0-inch schedule 40 type 30~ st inless di ch-="e piping of the "A" recirculation spray best exchanger a" Su~
No.
2.
A recent in"paction of Un't 1 Containment Recirculation Spray Piping revealed cracking s&ilsr to Unit 2.
On October 8, 1976, "nother incide " of similar cracking iz. 8-inch schedule 10 t3we 304 stainless piping of the S."fcty Infection Pi=p Suction, Line."t the Ginna faci'ity wa~ reported by the license Xn~orna"ion received an th" neta Iurgical analys.'s conducted to date ind" cares that the failures were the result
'o Intergr~aular tress corrosion cracMg th-t initiated cn the Inside of the piping.
A co~mality of actors oose~ed as oc~stcd vi.h the co rcsio" =ech"cim; were i The cracks were ad-acent to and p opag ted along weld zones of:hc thin-<<"alled lov. pressu e piping, not part of the reactor cool."t system 2.
Cracking occurred w piping containing re3atively stsgwnc, =ori=
acid solution ro-required for no~el operati g conditions.
Analysis of surface p oducts at this tim 'ndicate.
a chloride ion interaction m.th oxide fo~tion in the re3.atively sragnant boric acid solu=ion as the prob b3,e corrodant,
<<Cth the state of s"=ess probably due to ~aiding and/or fabrication The source of he chlorida ion is not definitely icno<<a.
bc+ever, AHO-I the chlorides and sulfide level observed 'n the surface tarnish 5"'>a near fields is*believed to have been introduced inta the p ping during esting of the soda hiosul ate discharge valves, o
valve leakage.
- Snarly, az GSn-a the chlorides and potenti-I cager,,
L"; Circular Ho. 7&46 November 26, l976 ava'"abiU.ty were assumed to have been present scca origina1 construction of tbe borated qatar "toraga tank which i" vented to atmo phare.
Corrosion attack at Surry is at~ributed to in-laaf~~a of ch3.orides through recirculation spray heat exchange
- tubing, a13,o ~ng bui1dup of contaminated wacar i" an othar~sa norual17 drv spra~ ~<nSns.
hCTX08 TO BZ ZKQ2'Y LICZRSFF.'-
provide a dasc iption o your program fo" assu ing continu<
integrity of those safety-related piping systems which ara not frequa t3y flushed, o: ~hich contain nc=lo~<ng 1iquids..This program should include consideration o
hydrostatic tasting in accordmxcc with ASM Code Section ZX ru3.es (1974 c.dition) fo all activa syst~
required for safety injection and containment
- spray, includS=g, their recircu3.ation modes, f om source af water supply up to the second isolation valve of the primary s)&tern, Solar test hould be cansiderad for other safety-ralat d pipin sy Your prop m hould also consider volumetric e:camination of a representative number o
~ circum; crential pipe walds by non-destructive mamination techn'ques.
Such ex~inations sho ld be'performed generally in accordance with hppand5x X of Section XX of the M.'1"- Code, except
=hat th examined area should cover a d=stance of approx~ately six (6) t~s the pipe vali thic~ass (but not )ess th n 2 Sncaas and naad noc exceed S 'aches}
on each side of the weld.
Supplemanta~
exa-ination techniques, such as radiography, should be usad where necessary for avaluation or confir=ation of u3.trasonic, indication" resulting fry such examination.
A report Baser bing your progran and schedule for chase inspec-t-ons should be subaittcd within 30 days fter rac ipt. of this CireDar.
The hRC Re3$.one Office should be'nfoaaed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of any adver a findings rasulti'ng during nondestructive evaluation of the acces ible pipw~g welds identi iad above.
A su~a~ report of the e"aainat'ons and evaluation of rasu3.t=
should ba submitted w.'th~ 60 days from the data of completion of proposed testing and examinations.
~
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lE Circu3.ar Ho. 76-06 3
Nova"her 26, 1976 This sw~ry report should 1so include a brief description of plant conditions, operating, procedures or other activities which provide assurance that the effluent chetd.stry Mill mintain lou leve3.s of potent~ corrodants in such relative1y s" g..ant regions vitMn the p'ping.
Your e ponses should be submitted to the Director of tMs office,
=wth a copy to the NRC Office of inspection and ">>".zorcment, Division of Reactor iaspec<<ion Propre--,
E~ashington, D.C.
20555.
Approval of %iC requ$ reaents for reports concerning possible generic problems has been obc ined under 44 U.S.C.3152 from the U
S General Account~my Office.
(GAO Approval 3-3.80255 (R0062}, expires 7/33.177. }
'IE Bulletin Ho. 79-17, Revision 1
Octobel 29, 1979 LISTI!iG OF IE BULLE, IIlS ISSUED IH LAST SIX tlOHTHS Enclosure Pace 1 of 3 Bulletin No.
Subject Date Issued Issued To 79-24 79-23 Frozen Lines Potential Failure of Emergency Diesel Generator Field Exciter Transformer 9/27/79 9/12/79 All power reactor fac i 1 ities which have either OLs or CPs and are in the late stage o: construction All Power '.eactor Facilities with an Oneratinq License or a construction oermit 79-14 Seismic Analyses For (Supnlement
- 2) As-Built Safety-Related Pipinc Systems 9/7/79 All Power Reactor Facil ities with an OL ora CP 79-22 Possible Leakage of Tubes 9/5/79 of Tritium Gas in Time-pieces for Lu...inosi y To Each Licensee who RecPives Tubes of Tritium Gas Used in Timepieces
'or Luminosity 79-13 (Rev.
1)
Cracking in Feedwater System Piping 8/30/79 All Desianated Applicants for OLs 79-02 Pipe Support Base Plate (Rev.
1)
Designs Using Concrete
'Supplement
- 1) Expansion Anchor Bolts 8/20/79 All oower Reactor Facilities with an OL or a CP 79-14 (Supplement) 79-21 79-2O 79 19 Seismic Analyses For As-Built Safety-Related Piping Systems Temperature Fffects on Level i~1easurements Packaging Low-Level Radi oacti ve ':!as te
. nr Transport and Burial Packaging Low-Level Radioactive
':.'aste for ransnort and ~v','al 8/15/79 8/13/79 8/10/79 8/10/79 All Power Reactor Facilities with an OL or a CP All PPIRs with an operating license All laterials Licensees who did not receive Bulletin Ho. 79-19 All Power and.,esearch Rpactol s with nLs,
'uel aci 1 ities
.;xco,nt
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.IE Bulle'in No: 79-17, Revision 1
October 29, 1979 Enclosure Paae 2 of 3 Bulletin No, Subject LISTINCi OF I E BULLETINS ISSUED IN LAST SIX t10NTHS Date Issued Issued To 79-18 79-GBC8(06C 79-17 Audibility Probl ems Encountered on Evacuation Nuclear Incident at Three
!file Island - Supp'iement Pioe Cracks in Stagnant Borated lfater Systems at P'; P, Pl ants 8/7/79 7/26/79 7/26/79 All Power Reactor Facilities with an Operatir!a License To'all PtlR Power Reactor Facilities with an OL All Pk'R's with operating license 79-16 Vital Area Access Controls 7/26/79 All Holders of and aoplicants for Power Reactor Operating Licenses who anticipate loading fuel prior to 1981 79-14 Seismic Analyses For (Revision 1)
As-Built Safety-Pelated Piping System 7/18/79 All Power Reactor Facilities with an OL or a CP 79-15 79-14 79-13 79-02 (Rev.
1)
Deep Draft Pumo Deficiencies Seismic Analyses for As-Built Safety-Related
'iping System C~acking In Feedwater System Piping Pioe Support Base Plate Designs Usina Concrete Fxpansion Anchor Bolts 7/11/79 7/2/79 6/25/79 6/21/79 All Power Reactor Licensees with a CP and/or OL All Power Reactor facilities with an OL or a CP All PNRs with an OL for action. All BMRs with a CP for information.
Al.l Power Peactor Facilities with an OL or a'CP
4 IE Bulletin No. 79-17, Revsion 1
October 29, 1979 Enclosure Page 'f 3 Bul'letin No.
Subject LISTING OF IE BULLFTINS ISSUED IN LAS: SIX HONTHS' Date Issued Issued To 79-01A Environmental Qualification 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental Qualification of ASCO Solenoid 'lalves)
All Power Reactor Facilities with an OL or CP 79-12 79-11 79-10 Short Period Scrams at BMR Facilities Faul ty Overcurrent Trip Device in Circuit Breakers for Engineered Safety Systems Reoualification Training Program Statist:cs 5/31/79 5/22/79 5/11/79 All GE BMR Facilities with an OL All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL
0 I
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