ML19242A744
| ML19242A744 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/26/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Short T OMAHA PUBLIC POWER DISTRICT |
| References | |
| NUDOCS 7908060128 | |
| Download: ML19242A744 (1) | |
Text
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Al; LING TON, T E X AS 76012 c
w July 26, 1979 Dochet No.
50-285 Omaha Public Power District ATIN:
T. E. Short, Acsistant General Mana;;ar 1623 liarrey Street Omaha, :.er;raska 63102 Centlemen:
The enclosed IC Dulletin 79-17 is forwarded to you for action. A written response is required.
If you deaire additional information regarding this matter, please contact this office.
Sincerely, 0
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Enclosures:
1.
IE Eullet'_n No. 79-17 2.
Listing, of IE Eulletins lusued in Last 12
!!onths cc:
R. L. Andrews, Manager Fort Calhoun Station Post Office Box 93 Fort Calhoun, Nebraska GS102 e/^7' i n
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l!NI'IED S ENfES NUCLEAR REGilLNFORY COMMISS10'i OlTICE OF INSPECTION AND ENFORCEMEfff WASH I NG'f 0N,
D. C 20555 I E Thil l e t i n !b. 79-11 Date July 26, 1979 Page 1 of 4 PIPE CR' IFS IN STAUNANr CORNIED WNIER SYSTEMS AT PWR PLANTS DescriN wa of Circsustances During the period of %vember 1974 t o February 1977 a niunber of cracking incidents have been experienced in safety-related stainless steel piping systems and portions of systems which contain oxygenated, etagnant or essent.i. ally s t agnant. ba ra ted wa ter.
Metallurgical investigations revealed these cracks occurred in the weld heat affected zone of 8-inch to 10-inch t ype 304 material (schedule 10 and 40), initiating oc the piping 1.D. surface and propagating in either an int ergranular or tran agranula r raode typical of Stress Corrosion Cracking.
Analysi: indicated the probable corrodents to be chloride and oxygen conta:nination in the a f f"cted systems.
Plants affected up to this time were Arkansa Nuclear Unit 1, R. E Ginna, II. H. Robinson l'ait 2, Crystal River Unit 3, San Onofre Unit 1, and Surty Units 1 and 2.
The NRC issued lE Ci rcular 76-06 (copy attached) in view of t he apparent generic nature of the problem.
During the refoc!!ng ontage of Three Mile Island Unit I which began in February of t his year, visual inspections disclosed five (5) threugh-wall cracks at welds
- i. the spent. f nel cooling syst em piping and one (1) at a weld in the decay heat removal system.
These cracks were found as a result of local boric acid build-up ad Jater confirmed by liquid lu'netrant tests This initial identification of cracking was reported to the NRC in a Licensee Event. Report (LER) dated !!ay 16, 1979.
A preliuinary metallurrital analysis was perf orrned by the licensee on a section of cracked and leaking weld joint irom the spent fuel cooling systen.
The conclusion of this analysis was that cracking was due to Intergranular St ress Corrosion Cracking (IGSCC) originating on tne pipe I.D.
The cracking was localized to the heat. a f f ected zone where the type 304 stainless steel is sensitized (precipitated carbides) during welding.
In addition to the main through-walI crack, incipient crack; were observed at several locations in the weld heat. affected zone including the weld root f ision area where a mini.scule lack of fusion had occurred.
The stresses responsible for cracking are believed te be primarily residual welding rtresses inasmuch a the calculated applied
- c. tresses were found to be less than code design limits There is no conclusive evidence at this time to identi fy those ag;;res sive chemical species which promoted this [GSCC attack. Furt her analytical ef forts in this area and on other system welds are being pursued.
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I' olletin No. 79-17 De July 26, 1979 Page 2 of 4 P,a d on tb: above analysis and visual leaks, the licensee initiated a b.
H isea ultrarsnic examination of potentially affected systems utilizing sr,, a tectniqws.
The systems exami.ned included the spent fuel, decay h ut r.Vtl,I skeup and puri fication, and s cactor building spray systems which ce t.it
,t agnant or int eroitteatly star,nant, oxygenated boric acid environ-2-1/J-inch (IIPSI) to ?.4-inch (borated c.e n t :.
me p tem range fr ri uiter
- rape ink suction), a:
type 304 itainless steel, schedule 160 t
to sci - ue 4o thicknes., respectively.
Inul t.s of t hese examinations were reporta4 to the LRC on June 30, 1979, as an update to the May 16, 1979 LER.
'the ultrasonic inspect. ion as of July 10, 1979, has identif ied 206 welds out of 94o inspected having UT indications characteristic of cracking randomly dist ributed throughout the aforementioned sizes (24"-14"-12"-10"-8"-2" etc.)
of the above :,ys t ems.
It is import. ant to note that six of the crack indicat ions were f oun ' in 2-1/2-inch diameter pipe of the high pressure inject. ion lines inside contatnment. These lines are attached to t he main cooiant pipe and are nonisoiable from the main coolant system except for check valves All of the six eracks u re found in two high pressure injection 1ines containing stagnated borated water.
No cracks were found in the high pressure injection lines which were occasionally flushed k -i ng makeup operat ions.
'l h e ultrasonic exaraination is continuing in order ts Q:iineate the extent. of the problen The above information was previously provided in In f orntion Notic" i r.
For All Pressurized Water Reactor Facilities with en Operating License 1.
Conduct. a review of safety-related stainless steel piping syster within 30 day, of the date of this lhilletin to identity 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 t ank (l',WST-RWST) piping.
(a)
Provide the extent and dates of the hydrotests, visual and volumetric c:caminations perf ormed per 10 CFR 50.55a(g) (Re: IE Circular 76-06 enclosed) of identified systems include a descript. ion of the non-destructive examination procedures, procedure qual i ficat. ions and acceptance criteria, the sampling plan, results of the examinations and any relate 1 corrective act. ions taken.
(h)
Provide :i description of wat"r chemistry cont.rols, staama ry of chemistry data, any design changes and/or actions taken, such as periodic flushing of reci rculation procedures,to maintain required water chemistry with respect to pH, B, CL, F, 0.
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3 IE 1;ulletin Ne. 79-17 Date July 26, 1979 Page 3 of 4 (c) Describe the pre.arvice NDE perf orud on the weld joints of identi.fied
- r. y s t e + >.
'J h e description in to include the applicable AT!E Code. ac-tions, d supplwent:, (.u!denda) that were foilowed, and the acceptance criterion.
(d)
Tacilitie, havinc, previously experienced cracking in identified wstems, l te;a 1, are re(pn:.;ted to identify (list) the new materialr utilized in repc.ir or replacement on a syst.em-by-system basit If a report of this information and that requested above has been previously subuit.ted to the NRC, please reference the specific report (s) in response to this Ilulletin.
2.
Fa c i l i t ie:. at whi.ch 1SI c:.aminations have not been performed (i.e., v i :.u ll and volumetric trf) on stagnant. portions of systems identified in Item 1 above, shJ 1 complete the fallowing actions at the earliest practical date but not later th;n 90 days aiter the dat e of the Unlletin.
(a)
Perfoco AS!!E Section XI visual examination (IWA 2210) of normally accessible welds of all engineered safety r,ystems at service pressure to verify system integrity.
(b) Conduct ultrasonic examination and liquid penet rant surface examination or a representative number of circumferential welds in nornally acce..sible' port ions of syste:as identified by Item 1 above It is intended that the sanple number of welds include all pi n? diam?ters i
in the 2-1/2 inch to 24-inch range with no less than a 10 percent nample by synten and pipe wall thickness It is also intended that the UT examination cover the weld fusion zone and a ninimum of 1/2-inch on each side of the weld at the pipe I.D.
The examination shall be in accordance with the provisions of AS5fB Code Section XI-Appendix III and Supplec ent <. o f t he 1975 Winter Addenda, except all signal responses shall be evaluated as to the nature of the indication 3.
These code methods or alternative examination met hods, combination of methods, or newly developed techniques may be used provide l the procedures yield a demonst rated ef fectiveness in detecting strest corrosion cracking in austenitic stainless : teel piping (c)
If cracking is identified during Item (a) and (b) examinations, all welds of safety-rclated piping systems and associated subsystems where dynamic flow conditions do not exist during normal operationa (Item 1) shall be subject to volumetric examination and repair including piping in areas which are normally inaccessible.
Norm. ally accessible.efers to those areas of the plant which can be catered x
during reactor operation.
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IE Bulletin No. 79-17 Date July 26, 1979 Page 4 of 4 3.
I d e a t. i f i ca t i on f craching in ocu-nni t. o f a r:ul ti-uni t f ac i l i ty wh i ch
, sa fet y-rellt ed nynt ems to be inoperable shall require inae d i a t e can-c
- aination of acce>.ible portions of other si:ailar unitz which have not.
e:
been inapected urah>r the ISI provinions of 10 CFR 50.55a(:;) unlesa justifi-ca t ' in for continuel operation is provided.
4.
A-T ra c k i n u, ident ified shall be reported to the Director of the a; pro-p:
itu ',RC Regior.al Of f ice 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 14 day scritten report.
a 5.
Prov2de a written report to the Director of the appropriate hRC Regional Office within 30 days of the date of this Bulletin addres;ing the results of your review required by Ite:a 1, 6.
Complete the o mmination required by Item 2 within 90 days of the date of t hi., Bulletin and provide a wri tten report to the Director of t he appro-priate NRC Regional Office within 120 days of the dat e of this Bulletin dencribing the results of the inspections required by lten 2 and any corrective r;"asures taken.
7.
Copies of t he reporta required by Iteus 4, 5 and 6 above shall be provided to the Director, Division of Operating Reactors, Office of Inspection and Enforcement, Washington, D. C.
20555.
Approved by Gel, nl80225 (I:0072), clearance expires 7/31/80.
Approval wa:, :;iven under a blanket clearance specifically for idenified generic problems 1!A
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I!: Circu;lar 76-05 1:oveeder 24,19/6 STESS CO:UtOST.Or. Ci'aci:S It! ST.'.CD' T, LO'! PP.1:S."U:lE STAl?!LESS PIPII;G CO::T?sD;ING EO: TIC ACID SOLUTIO:: AT Pun'-
IE ; Z_IPTION O'c CIR CC:4ST!':ri:S :
Ik.r: L th: p a r i o d ' o m.b n 7, 19 74, to I:ove: Sr 1,1915, neveral inci-Cant: O E throug'.-va'.l cracking hava occurred in the 10-inch, schrdulu 10.p = 304 stainlaan nc'el piping of the Reactor nullding Spray and Deca /.ient Synt.<.O at Arkanua;, I;uclear Plaat 1;o. 1.
Oa October 7,1976, Virginia 1:lectric and Power also reported throagh-uall cracking in the 10-inch neh, dule 40 type 305 stainleu, discharge pipin3 of th "A" recirculation npray heat e::chan;;er at Surry Unit I!o. 2.
A receat inspection of Unit Ilo. 1 Containnwnt Recirculation Spra) Piping revealed cracking sinilar to Unit 1:o.
2.
On October 8,1976, another incident of siallar crackina :in 8-inch sched-ule 10 type 304 ntainless piping of the Safety Inject loc. Pung Suction l.ine at the Cinna facility um. reported by the licennee.
Information received on th-. at allurgical analysir, conducted to date indicates that.the fai. lures were the rouult of intergtanular stret.a I
corronion crachinathat initiat.d on t he innide of thr-piping.
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co:2 anality of f actorn ob:,erved associat ed uith th' corronica mechaninu were:
1.
The crach,.:cre cdjacent to and propay ited alona weld nes of the thin-valled les pressure piping, not part of the rene
- coolant system.
2.
Cracking occurred in piping containing, relut. eciy rearnaut boric acid r.olution not required for nornal operating conditions.
3.
Analynir of nurface products at thi tiu indicate a chloride ion interaction wi th o:<ide for.mation in the relativc-ly nta".nant horic acid no]ution.
the probable corrodant, wi th the nt.-
e of strenn probably due to welding and/or f abricatica.
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.>urce of t he chloride ion Jr. not def init ely huoun.
IIouever, at Ih' the cMorld:. and nulfide i< vel observed in t he nurface tarnit.h f i '. m near u 1d'. In believed to have been int roduced into t he piping ou r'
- tentiur oi t '.u. nod ina t.h ioau1E ate d incharge valvet., or valve.
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Sini.larly, at Giuna the chloridea and potent ial o>:yy.a wrail-t
_..y were 2 2 :cl t. 's Eve Lee, pre:. nt c inci orin i.nal cona truction c.
- hora ted ua te r storage ta.
vhi.ch is vented to atuouphere. Corro-
- .n.. _Ltack at Surry is attributed to Ju -leakaa. of chlorld'. through recirculatica ': pray heat e;.: change tubing, alloui.ug builduj of contaminated water in a a oti, ervine earnally dry spray piping.
ACT7.0'I TO BF. TAmi BY 1,1CliNSEE:
1.
l'ro vid e - a dencription of your prograra for anauri.n; continued integri.ty of those nafety-related pip.i.ua syntena ubich are not frequently fluuhed, or which contain nonflouing liquida.
Thir progra:a chould.includa cotr-t.ideratio: of hydroatatic t e n t i u ". In accordance with ASm Code Section-(1974 Edit ica) for all act ive nynt"na re.luired for safety XI rule i nj ec t ion and containmut. spray, inclu ling the i.r rec irculation nodes, from nource of u.it er :.upply up to the necond isolation valve of the prinary yate.a.
Slallar t ents nhould be ecnuidered fot other na Enty-rel ated pi pj ag :.yntenn.
2.
Your pro;p'au nhould alno concidor mlum at ric c: uinat ion o f.. <epre-mber of circunterenc ial nipe vold:. by nond:. t ru c tivt n e n t a t l'. t-n:
exami nat. i on techn iane...
Such enaninat. ions should be p"rf ormed generally la accordance uf t h Append h 1 of Section XE ot the AS:@.
Code, except t.h a t the examined.> : ea should cov ar e distance of approxi.-
nately six (6) tir,. the pipe val? th ichnr.-
(but not le;,<
than 2 iecht and eeed not exceed 8 Inches) oa each nide of the veld.
Supple:r ata ry exaainat i.on techn iques, such as radiography, c'
- ld be used there necen<ary fcr evaluati.on or confirmation ot ul l-
.onIt ind icat ions renulting fron nuch exaninat ion.
3.
A report describing your progran and schedule for t.hecie ' inspections should be nubmitted uithin 30 day: after receipt o f: thiu circular.
4.
The 1:P.C negion.'.1 Office chauld be I.. corned uithin 24 hoorn, of any adverse findings renulting during nondestructive evaluation ef t.he accennihle piping ueldn ideatifled above 5.
A nu mary report of the exaninationa and evaluation of re.ultu nhould be subnitted within 60 day. f roa the date of completion of proponerl testing and examinations.
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, This nurmary report should also include a brief deraription of plant condicions, operatina proceduren or other activities which providi assurnace that the effluent cheaintry will mtintalu low levels of potential corrodante in such relatively stagnant regi.ons within tite piping.
You responses should be c.ub;titted. to the Director of this of Eice, with a copy t, the l'RC Of fice of Inspectica and Enforcenant, Division of Iteac tor.tnsp actica Progra u, Washington, D.C. 20555.
Approval of I!RC requireaants for reports concerning possible generic probleas has been obtained under 44 U.S.C. 3152 from the U.S. General Accounting Of fice:.
(CAO Approval B-1802SL (R0052), expires 7/31/77).
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IE Bulletin ?;o. 79-17 July 26, 1979 LISTING OF IL till LETIt'r ISSUED IN LAST TWblVE MON uS I;ull e t i a Subject Date issued Issued To No.
7.-11 Examination of Mark 1 7/21/78 BWR Power Reactor Contain:*nt Torus Fa c i l i t i e:, for Welds action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hat ch 1, !!ont.i-cello and termont Yankee 78-12 Atyp; al Weld flat eriel 9/26/78 All Power Reactor in I!eactor Pressure Facilitie:, with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP)78-12A Atypical Weld flat erial 11/24/78 All l'ower I cactor
- i. Icattor Pressure Facilities with an el Welds Operating License (OL) or Construc-tion Permit (CP)78-12B Atypical Weld !!aterial 3/19/79 All Power Reactor in Reactor Presrice Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit. (CP) 78-13 Failures in Source 10/27/78 All General and lleads o f Kay-Ray, Speci fic Licenseer, Inc., Ganges !!odels with Kay-Ray Gauges 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 70611; Enclosure Page 1 of 4
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U' Ihilletin No. 79-17 July 26, 1979 78-14 Deterioration of 12/19/78 All GE nWR facilities Buna-N Compon>nts uit h an Operating in sCO Solenoids License (OL) or Construction Permit (CP) 79-0!
Envi ronniental Quali-2/8/79 All Power Reactor fication of Class IE Facilities with an Equipment Operating License (OL) or Constructica Permit (CP) 79-Ola Environmental Qualification 6/6/79 All Power Reactor of Class IE Equipw nt Pacilities with an Operating License (OL) or Construction Permit (Cl )
79-02 Pipe Support Base 3/8/D All Power Reactor Plate Designs Usiny Facilities with an Concrete Expansion Operating License Anchor Dolts (OL) or Construction Permit (CP) 79-02 Pipe Sul, port Da:;e 6/21/79 All Power Reactor (Rev. 1)
Plate Designs Using Facili. tic, with an Concret e E>:pansion Operatin License Anchor Bolts (OL) or Lonstruction Pe rrrit (CP) 79-03 Longitudinal Ucid 3/12/79 All Power Reactor Defects In ASME SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spools (OL) or Construction flanuf actured by Permit (CP)
Youngstown Welding and Engineering Company 79-04 Incorrec t Wei,,htr.
3/30/79 All Power Reactor for Swing Check F:.cilities with au Valver Man actured Operating License by Velan Engineering (9L) or Cones ;,etion Corporation l'e t m i t (CP)
Enclosure Pa n' 2 o c 4 tj O ',/b
- 7) )t r)
IE Ilullet in No. 79-17 Jui; 26, 1979 79-05 Nuclear Incident at 4/1/79 All Power P. ctor
'I n ree ?!i l - Island Facilities with an Op
- Ling License (O.. or Construction Pe iit (C P).79-05A Nuclear Inci ent at 4/5/79 All Power t actor Three flile,aland Facilitia ith an OperaLin<
!. : cense (3L) or Coastruction Permit (CP)79-05B Nuclear Incident at 4/21/79 All 1:5W Power Reactor Three tiile Island Facilities with an Operating License (OL) 79-06 Review of Operational 4/11/79 All Pressurized K.. tar Errors and System Power Reactor racilities IIisali.guments Identifie l Except I)&W F ilities During The Three Ilile Island Incident 79-06A Review of Operational 4/16/79 All Westinghouse PWR Errors and System Facilities wit h an IIisaligni:ents Identified Operating License During the Three !!ile (OL)
Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev. 1)
Errors and Sy ; tem !!is-Power Reactor Facili. ties alignments Identified of Westinghouse Design During the Three tiile with in Operat.ing License (OL)
Island Incident 79-06f; Review of Operational 4/14/79 All Combustion Engineer-Errots and Svstem ing PWR Facilities with
!!i sa l i gtunents Identified an Operating License During The Three tiile (OL)
Island 79-07 Seismic Stress Am lysis 4/14/79 All Power Reactor of Safety-Rela Rd Fiping l'acilities ui.th an Operating License (OL) or Construction Permit (CP)
Enclosure Page 3 of 4
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IE IDillet in ;a. 79-17 July 26, 1979 79-08 Events Relevant to UWR 4/14/79 All BWR Power Reactor Reactora Identified Fa c i 1 it i e.,
witb
- i During Three Ilile Island Operating Licen.>
Incident (OL) or Construction Permit (CP) 79-49 Fa i. l u r e: of t ' Type AK-2 5/11/79 All Power Reactor Circuit Dreaker in Sately Facilitie:. with an
% Lated Syst, as Ope ra t. i ng '.icense (OL) or Constructice Permit (CP) 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an nperating License (OL) 79-11 Faulty Overcurrent Trip 5/22/79 Al'1 Power Reactor Device in Circuit Breakers Facilities 6 th an for Engineered Safety Jperating License (OL) or S. :;tems a Const.ruction Permi.L (CP) 79-12 Short Period Scrams at 5/31/79 All Power Reactor Facilities BUR Faci.lities with an Operatir.g License (OL) or a Construction e nn i t (CP) 79-13 Craching In Fec< heater All PWRs wit h an Operat ing Syntem Piping License (OL) for action.
All DWR with a Co n s t.ru c t i o n Permit (CP) for information
,9-14 Seismic Analyses for 7/2//9 All Power Reactor facilities As-huilt Safety-Related with an Opertin;; License Piping Sys'.cm (OL) or a Construction Permit (LP) 79-15 Deep Dratt Pump 7/11/79 All Power Reactor Facilities Deficiencie:,
with a Construction Permit and/ar Operating License (OL) 79-16 Vital Area Acce:, Controls 7/26/79 AlL Power R a ctors w.
- i an Operating License (OL) or anticipating fuel loading, prior to January 1981.
Eaclosure Page 4 or 4 hk\\
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