ML19247B121
| ML19247B121 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/26/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Proffitt W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 7908070771 | |
| Download: ML19247B121 (2) | |
Text
-
+
Mpm a r c 9o UNITED STATES
/
l i
NUCLEAR REGULATORY COMMISSION I?
'^r REGION 11 I,.
,h,;
[
AT L ANT A. GEORGI A 30303 101 MARIETT A ST, N W, SUIT E 3100 3*
g '.... a N
E' In Reply Refer To:
RII:JP0 50-339 50-404 50-405 Virginia Electric and Power Company ATTN:
b'. L. Proffitt Senior Vice President, Power P. C.
Box 26666 Richmond, VA 23261 Gentlemen:
No The enclosed Bulletin 79-17 is forwarded to you for infor ation.
written response is required. However, the potential corrosion beha-vior of safety related systems as it regards your plant over the long ter should be taken into consideration.
If you desire additional inf ormation concerning this catter, please contact this office.
S inc.er ely,
_r.
y (b.,/
-x 41 Jaces P. O'Reilly Director Encl;sure:
IE Bulletin No. 79-17 c p l.
l_ ()
o Jv 7908070771 e
q
Virginia Electric. and p.
Power Coorpany >
ec w/ encl:
W. R. Cartwright, Staticn Manager Post Office Box 402 g,
Mineral, Virginia 23117 g,
P. G. Perry I'
Senior Resident Engiree:
f Post Office Box 38 f
Mineral, Virginia 23117 W.
L. Stewart, Manager Post Office Box 315 g.
Surry, Virginia 23883 h.. : 3 F
t l
b.a
.M t4L fl p
i 4W pi:
f i
i
?cli. > Yk'
!0 i
k
,k w
a th 6
,9 T.i.
7 7 6 L... % h,....,, % , t.--.1 4 6..i
~ ' - ~.
143..'.._ - t.
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C.
20555 July 26, 1979 IE Bulletta No. 79-17 PIPE CRACKS IN ST AGN ANT 50 RATED b ATER SYSTEMS AT PWE Descripticn ;1 Circamstances During the period of Ncvember 19'a to February 1977 a number of crack; safety-related stainless steel piping systers and or essenttall'. stagnant experienced in have h en stagnant systems which conta:n oxygenated, revealed these cracks accurred ;n porti:n
- c. f Metallerg cal investigat:ons 304 mater:al ;schedZe berate 3.ater.
10-inch type affected :ane of S-inch to en the p; ping !.D. surface and propagating in either an the weld heat transgranular made typical of Stress Ccrrcsicn Cracking-and
-C',
initiating corrodents to be chl r:de and cxygen c;ntar:-
i r.t e r g r a r.u l a r or Ar.. c s i s :nd:cated the prctable Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Giana, H.B. Ecbinson Unit 2. Crystal River Unit 3, San nat:cn in the affected systems.
1, and Surry Units I and 2 The NRC issued Circular 76-06,ccpy in view of the epparent generic nature cf the problem.
Oncfre Unit attached)
Island Un:t I which began in Februan refueling outage of Three Mile through-wall cracks at welds Dur:ng the of this year, visual inspections disclosed five (5) a weld in the decay heat fuel cooling system pip:ng and one (1) of local boric acid build-at in the spent These cracks were found as a result Inis initial ident;ficaticn remc-cal system.
tests.
ccnfirmed by liquid penetrant (LER ) dated M23 reported to the NRC in a L:censee Event Report up and later licensee en A preliminary metallurgical analysis was perf ormed by the of cracking was from the spent fuel cooling system.
lt, l'7i.
a sect:an of cracked and leaking weld joint The ccc:!usicn of this analysis was tb't cracking was due to Intergranular (IGSCC) originating on the pipe !.D.
The crack:ng stainless steel :s Stress Carrcsion Cracking af f ected zone where the type 304 senstt. zed (precip:tated carbides) during welding.
In addition to the ma n was 1: cal: zed to the heat through-wall crack, inc:ptent cracks were observed at several locations in fusion area where a mir:scule we d heat affected zone including the weld root for cracking are believed responsible lack cf fus 'n had occurred. The stresses in as much as the calculated appl.ed residual welding stresses less than ccde des 2gn limits. There is no conclusive te be prima.
y h prcmeted stresses were fcund tc be evidence at this time to identif y those aggressive chemical spe this IGSCC attack.
welds are being pursued-Fi',
n,3 O
L.
[j j 7908020183
July 26, 1979 IE Bulletin No. 79-17 Page 2 of 4 inttrated a broad leaks, the licensee Based on the above analysts and visualexamination of potentially af f ected sys al based ultrasonic examined included the spent fuel, decay beat removal, techniques.
The systems which contain stagnant makeup and purtf tcation, and reactor building spray systems These systems acid environments.
intermittently stagnant, oxygenated boric tank sucticni, range from 2 1/2-inch (HPCI) to 24-inch (borated water storage hickness respectively or are type 304 stainless steel, schedule 100 to ;chedule 40 t the NRC cn June 30, 1979 as an reported to tnspection as of July
- 10. 1979 Results of these examinations were update to the May 16, 1979 LER. The ultrascnic harac stt:
has identified 206 welds out of 946 inspected having UT indications caforemention l-'-
the cracking randomly distributed throughout to note tha. six of 12"- 10"-S"-2" e t c. ) of the above systems.
It ts important ot r pipe cf the h:gh pressure t
the crack indications were found in 2 1/2-inch diame eThese Itnes are at for check calves.
inje iton 1:ces inside containment.from the main coolant system except pipe and are acnisclable lines containing All of the six cracks were fcund in two high pressure injection the high pressure inject:cn No cracks were found in The ultrascct:
staznated bcrated water.
lines which were occas:enally flushed during makeup operations.f the problem.
examination is continuing in order to delineate the extent o The abcve inf ormation was prev:ously provided in Inf ormation Notice 7 For All Pressurized Water Reactor Facilttles with an Operating License:
30 of safety related reainless steel piping systems withinid 1.
Ccnduct a review days of the date of this Bulletin to ubich contain stagnant oxygenated borated water. These systems l, spent fuel pool cooling, typically include ECCS, decay / residual heat removaccntain systems Frovide the extent and dates of the hydrotests, visual and volumetric 10 CFR 50.55a(g) (Re: IE Circular 70-06 (a) examinations performed per description of the non-Include a encicsed) cf identified systems.
and destructive examination procedures, procedure qualtftcations criteria, the sampling plan, results of the examinat:cns acceptance and any related corrective actions taken.
Provide a description of water chemistry control (b) ith rectrculation prgcedures to maintain required water chemistry w ofrespect to pH, B, CL, F, 0 '
2
. - =
July 26,1979 IE Bulletin No. 79-17 Page 3 of 4 f toentitled Describe the preservice NDE performed on the weld sec-(c) systems.
tions and supplements (addenda) that were criterion.
d systems,
Facilities having previously experienced cracking in identifiethe ne Item 1, are requested to identify (list)
(d)
If a report cf system-by-system basis.
requested above has been previcusly submitted or replacement on a repair tc this information and that to the NRC, please reference the speclitc reportts) in respcnse this Bulletin.
3 (1 e., 'isual Facilities at which ISI examinations have not been perfcrme identified in Item I UT ) on stagnant portions of systems the earliest practical date 2.
and volumetric shall complete the following actions at but not later than 90 days af ter the date of the Bulletin.
- above, XI visual examination (I'A 2210) cf no rmal 3 t service pressure Perform ASKE Section accessible = welds of all engineered safety systems a (a) to verify system integrity.
tion Conduct ultrasonic examination and liquid penetrant surf ace exam in normally acces-or a representative number of circumferential welds (b)
It is intended that sible" portions of systems identified by 1 above.all pipe diameters in the 2-1.2 include sample number of welds sample by system inch to 24-inch range with no less than a 10 percentis also intended the It h en each side of and pipe wall thickness.
the weld f usion zone and a minimum of 1/ 2-inc cover the pipe I.D.
Appendix III and Supp'.ecents the weld at the provisions of ASMI Code Secticn XI -of the 1975 W hall be evaluated as to the nature of the indications.
l developed techniques examination methods, combination of methods, or new y ted effectiveness may be used provided the procedures yield a demonstraaustentric stain corrcsion cracking in in detecting stress piping.
all and (b) examinaticas, identif ted during Item (a)d associated subsystems 'here If cracking is welds of safety-related piping systems anduring normal operaticas (c) iltem 1) exist ng dynamic flow conditions do notto volumetric examinatten and repair including shall be subject in areas which are normally inaccessible.
be entered
- Sormally accessible refers to those areas of the plant which can during reactor operation.
C.o fl )
July 26,1979 IE Bulletin No. 79-17 Page 4 of 4 3
Identification cf cracking in one unit of a multi-unit f act11ty which safety-related systems to be inoperable shall require treediate of other similar units which have not causes ISI provisions of 10 CFR 50.55a(g) unless justift-portions examination of access ble been inspected under the cation f or continued operation is provided.
Any crack:ng identif ted shall be reported, o the Director of the apprcpriat 14 day 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 identiftcation followed bv a 4.
written report.
the Director of the appropriate Nhc Regional results to 5
Previde a written reportof the date of this Bulletin addressing tt.e Office within 30 days of your review required by Item 1.
f examination required by Item 2 within 90 days of the date c f the appropr: ate the Complete this Bulletin and provide a written report to the Director oNRC 6
Iter 2 and any correct:a measures the results of the inspections required b) taken.
Cop:es of the reports required by Items 4, 5 and 6 above shall also be Inspec-provided to the Director, Dtvision of Operating Reactors, Office of 7.
20555 tion and Enf orcement, Washington, D.C.
clearance expires 7/31/80. Approval was given Approved by GAO, B180225 (R0072),under a blanket clearance specifica blems.
Enclosures:
1.
List of IE Bulletins Issued in 1979
- a w
(
L,
November 26, 1976 II Circular No. 76-06 URI STAINLISS STP2SS CORROS10$ ~ RACKS IN S DESCRIPTION OF CIRCQ'.STECES:
1,1975, several incide::s Durirg the period Nove=her 7,1974 to Nove=berin the IO-inch, sr.hedule 10 rype of through-vall cracking nave occurred 304 stainless Spray and De:sy Eest No. 1 Arka sas Nuclear Plan:
Remov al Sy s t et.s at ise reperted :Fr: ugh-On October 7,1976, Virg1. cia Electric and Power a stainless dischstge 40 type 30' vall cracking in the 10-inch schedule exchanger at Surry U;i t piping of the "A" recirculation spray heat Recirculatice Spray 1 Containment inspection c f Ur.it 2.
No.
2.
A recent Piping revealed crackAng si=ilar to Unit of similar cracking in 6-inch Sr f cty inj ectie: Pu:p On 0::cber 8,1976, another incidents::1 less piring of the the licensee.
schedule 10 type 304 facili:y was reper:ed by the Cinta Suetice Lice st conducted to da:e l
i Inf orsatien received on the netallurgical ana ys rof intergicnv'2r etress the resul:
A the failures wereinitiated on the inside of the pipinl.i h the indicates tha:
corrosion cracking thatcc=menali:y ef f ac:crs cbserved ascociated f rne were:
to and propagated aleng ucid :cces oef the re The cracks were adjacent r:
thin-valled lov pressure. piping, not pa 1.
l agnant boric system.
Cracking occurred in piping centaining relative y stre 2.
acid selu: ice cc:
this time indica:e a chler de ics l
fvaly stagnant beric Analysis of surface products atinteraction vith c:1de 3
probeble corredant, acid solution as the io n.
probably due to velding and/or f abricat Novever, c:
dafini:cly knovn.
d in the surface tarnish is not The source of the chloride iot AND-1 the chlorides and sulfide level ebserve d into the pipteg film near velds is believed to have been introducesodium t during testing of the the chlorides and potential oryge:
Cin:a Si=ilarly, at Irakage.
E10 m,-
d/j v-O
Novecher 25, 1976 2-II Circular No. 76-06 since original t
availability were assumed to have been presenk which is vaated to construction of the borated water storage tanSurry is attributed to in-leakage of Corrosion attack at exchange tubing, alleving atmosphere.
l ion spray heat lly dry sprsy piping.
, chlorides through recircu atbui16up of contam:ina:ed water in an oth ACTICS TO BE TAKDi ET LICENSEE:
i ed Provide a description of your progra= for assuring cont nu tich are not integrity of those safery-related pipsag systems v This ic :esting in progran should include consideration of hydrostat (1974 Edition) for accordance with ASMI Code section XI rules ion and containment all active syste=s required for saf ety injectfrom source of water includi=g their recirculation mOder, i
y sys te:.
supply up to the second isolation valve of the pr
- spray, l
d p: pin; systems.
ico of a Your progra= should also consider volum4:cie exacina:
lds by non-representative number of circumf erential pipe weSuch exa:inatiens sho 2.
destructive examination techniques.be performed gener that the examined area Section XI of the ASMI Code, exceptshould cover a d
) times the (but not of the veld.
Sup;1esee:ary pipe wall thickness exceed S inches) on each sidt ld be used exasination techniques, such as radiography, shou f ultrasonic where necessary for evaluation or confirrstion o such examination.
indications resul:ing f ro:
and schedule for these inspec-3escribing your progra:
fter receip: of this A report tic:s should be sub=1:ted within 30 days a 3
Circular.
The NRC Regional Of fice should be informed within '24 bour ie of any adverse findings resulting during nondestruct v 4.
d ified above.
evaluation of the accessible piping velds i 2nt of the exa=inationa and evalus:1on of results f completica A summa:y reportshould be submitted within 60 days f rom the date o 5
of proposed testing and examinations.
O O
e
.s November 26, 1976
' II circular No. 76-06 i
of sbould also include a brief descript on ctivitias This stmaary reportconditions, operat: Lng proceduras or other achemist the affluent i ly stagt. ant plant ubich provide assurance thatlov levels of potential corroda f
regions within the piping.
tor of this office, i
Your responses should be submitted to the D reca2.d Enforcement Divis i
with a copy to the NRC Cf fice of Inspect onInspection Progr 20555 of Reactet erning possible geeets.
Approval of h*RC requirc=ents f or reports concC 3152 f rom the U.S. G expir es 7 / 31/ 77. )
problems has been obtained under 44 L S.(GAO Approval B-1802 Accounting Of fice.
e 9
e
'P W
[
l Jb c,
.f' T
O O
~
.y
Enclosure :
IE Bulletin No. 79-17 Page 1 of 4 July 26, 1979 LISTING OF IE BULLETINS ISSED IN LAST T4IXE MONTHS Date Issued Issued To Bulletin Subject No.
All Holders of and 7/26/ 9 7
Vital Area Access Controls applicants for Reacter 79-16 Operating Licenses All Power Reactor 7/11/79 79-15 Deep Draft Pump Licensees with a CP Deficiencies and/or OL All Power Reactor 6/2/79 Seismic Acalyses for facilities with an 79-le As-Built Safety-Related OL or a CP Piping P stem All PWRs with an 6/25/7)
Cracking in Feedwater OL for action. All 79-13 System Piptng BWRs with a CP for infor=ation.
All Power Reactor 6/21/79 Pipe Support Base Plate Facilities with an 79-02 (Rev. 1)
Designs Using Concrete OL or a CP Expansion Anchor Bolts All GE BWR Facilities 5/31/79 Short Period Scrams at with an OL 79-12 BWR Facilities 5/22/79 All Power Reacter Faulty Overcurrent Trip Facilities with an 79-!!
Device in Circuit Breakers OL or a CP for Engineered Safety Systems 5/11/79 All Power Reacter 79-10 Requalification Training Facil:ttes with an OL Program Statistics 4/17/79 All Power Reactor Failures of GE Type AK-2 Facilities with an 79-09 Circuit Breaker in Safety OL or CP Related Systems All BWR Power Reactor 79-08 Events Relevant to BWR 4/14/79 Facilities with an OL Reactors identified During Three Mile island Incident
,a
- \\ 's Gc O
IE Bulletin rio. 79-17 Page 2 of 4 July 26, 1979 LISTING OF IE BL'LLETINS ISSLTD IN LAST TVEIXE MONTHS Date Issued Issued To Bu11etta Sub;ect No.
79-07 Seismic Stress Analysis 4/14/79 All Power Reacter Facilities with an of Safety-Related Piping OL or CP All Combustloc Engineer-79-06B Review of Ope-ational 4/14/19 ing Designed Presscr; zed Errors and System Mis-Wate Power Reactor alignments Identified Facilities with an During the Three Mile Operating License Island Incident 4/15/79 All Pressurized Water 79-36A Review of Operational Power Reactor Faci.. ties (Re/ 1)
Errors and System Mrs-of Westinghouse Design alignmunts Identified with an OL During the Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Pressurized '- ter Power Reactor Factiltics Errors and System Mis-of Westinghouse Eesign alignments Identified with an OL During the Three Mile Island Incident 79-06 Review of Operat.onal 4/11/79 All Pressurized Water Power Reactors with an Errors and System Mis-OL except B&W facitities alignments Identified During the Three Mile Island Incident 79-05A Nuclear Incident at
~/5/79 All B&W Power Reacter Faci.ities with an OL Three Mile Island 4/2/79 All Power Reactor 79-05 Nuclear Incident at Facilities with an Three Mile Island OL and CP
Entiosure 2 IE Bulletin No. 79-17 Page 3 of 4 July 26, 1979 LISTING OF IE BULLETINS ISSUED IN LAST T=IlXE. MONTHS Date Issued Issued To Bulletin Subject NO.
79-04 Incorrect beights for 3/30/79 All Power Peactor facilities witt or Swing Check Valves OL or CP Manufactured by Velan Engineering Corpcration 78-12B Atypical Weld "ater tal 3/19/79 ul Power Reactcr facilities with an in Reactor Pressure OL or CP Vessel 'aelds 79-03 Lccgitudinal Welds Defects je12/79 All Power Reactcr Facilities with an In ASME SA-312 Type 30' OL or CP Stainless Steel Pir-Spools Manufactured b) iaungstown Welding ar.; Engineering Co.
79-02 Fipe Support Base Plate 3/2/70 All Power Reacter Facilities with an Designs L' sing Concrete OL or CP Expansion Anchnr Bolts Environmental Qualification 6/6/79 All Power Reactor 79-OlA Facilities with an of Class IE Equipment OL or CP (Deficiencies in the Envi-ronmental Qualificaticn of ASCO Solenoid Valves) 79-C1 Environ = ental Qualification 2/8/79 All Power Reacter Facilities with an of Class IE Equipment OL or CP 8-la Deterioration of Buna-N 12/19/78 All GE BW7 fact.ities with an OL or CP Ccepcrent In ASCO Solencids r,(,',
?'O JdL
/dV
IE Bulletin Nc. 79-17 Page 4 of 4 July 26, 1979 LISTING OF IE BL'LLETINS ISSUED IN LAST TVELVE MONTHS Date Issc.d Issued To Bulletin Subject No.
Failures in Source Heads 10/27/78 All general and specific licensees 78-13 of Kay-Ray, Inc., Gauges with the subject Models 7050, 7050B, 7051, Kay-Ray, Inc.
7051B, 7060, 7060B, 7061 gauges and 7061B 7S-12A Atypical Weld Material 11/24/78 All Power Reactor Facilities with an in Reactor Pressure OL or CP Vessel Welds 9/29/78 All Power Reactor Atypical Weld Materia.'
Facilities with an 75 in Reactor Pressure OL or CP Vessel Welds Examination of Mark I 7/21/78 BWR Power Reactor Facilities for action:
TS-11 Containment Torus Welds Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee cn,
,a1 Jw iJl
.