ML16341B234
| ML16341B234 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/29/1979 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Crane P PACIFIC GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7911150469 | |
| Download: ML16341B234 (24) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V 1990 N. CALIFORNIA8OULEVARD SUITE 202, WALNUTCREEK PLAZA WALNUTCREEK, CALIFORNIA94598 Octob r 29, 1979
~C. V Docket I'los. 50-275 50-323 Pacific Gas and Electric Comoany 77 Scale Street San Francisco, California 94106 Attention:
t1r. Philip A. Crane. Jr.
Assistant General Counsel Gentlemen:
The enclosed Bulletin Ho. 79-17, Revision 1 is forwarded to you for information.
No written response is required.
However, the potential corrosion behavior of safety-related systems as it regards your plant over the long term should be taken into consideration.
Ef you desire additional information concerning this matter, please contact this oTfice.
Sincerely, R.
H. Engelken Director Enclosure; IE 8ulletin Ho. 79-17, Revision 1
cc w/enclosure:
J. Morthington, PGSE M. Raymond, PGEE R.
- Ramsay, PGSE, Diablo Canyon
UNITED STATES SSIHS No.:
6820 NUCLEAR REGULATORY COD!11SS ION Accession Ho.:
OFFICE OF IHSPECTIOH 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 Description of Circumstances:
IE Bulletin No. 79-17, issued July 26, 1979, provided information on the cracking R1 experienced to date in safety-related stainless steel piping systems at PWR.
Rl plants.
Certain actions were required of all PWR facilities with an operating Rl 1 icense within a speci fied 90-day time frame.
Rl After several discussions with licensee owner group representatives and inspection Rl agencies it has been determined that the requirements of Item 2, particularly Rl the ultrasonic examination, may be impractical because of unavailability of R1 qualified personnel in certain cases to complete the inspections within the tim Rl 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 R1 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-qted water.
t<etallurgical 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,'an Onofre Unit 1, and Surry Units 1 and 2.
The NRC issued Circular No. 76-06 (copy enclosed) in view of the apparent generic nature of the problem.
During the refueling outage of Three tlile 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 orie (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'dentification of cracking was reported to the NRC in a Licensee Event Report (LER) dated i~lay 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.
R1
- Iden:i."es
'.hose additions or revi i"n
".o IE Bulletin i 0 ~
79-17
T
')
IE Bulletin No. 79-17 Revi s ion 1
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 R1 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 facile Island Unit 1 indicates that the extent R1 of IGSCC experienced in stainless steel piping at that facility may be more Rl
'imited than originally stated above.
Of the 1902 total welds originally Rl inspected 350 contained U.T. indications which required further evaluation.
R1 These 350 welds have been reinspected with a second U.T.'rocedure which pur-Rl portedly provides better discrimination between actual cracks and geometrical R1 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
I I
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~ hler 4et cleara>>r~ srocifqcal1v fo) idept':<ied
~c',".~! ic p;.! r<<4lewc.
Frc 'osv'"~s:
l.
IE Circular 'Io. 76-AG 2.
List o T" tulle ins Issued in
".!;-.. L'..s-. Six '!onths
~ ~
Hov~ber 26, 1976 IE Circular Ho. 7&46 STB:-SS CORROSION CR'CM l'H STACRAh, LQV PRZS KK STEZNIZSS PZPXN CGH A".'~uXG 30PZC ACID SOLUTXOP AT HQ's DZSc. GPTXOP C
CIRCRf5~CZS
'uring
<<he period 'Boveaber 7, "974 ta November 1, 1975,seve-al incidents af through-wall cracking h~~e occur ed in the lo-inch, schedule 10 type 304 stainless stee3. pipe@ of the Reactor BuVding Spray and Decay B~t Re=ova3 Sys ~ at Arkansas boucle'P3.ant Ho. l.
On October 7, 1976, Virginia,Zlectric and power also reported
=hrazgh-
'~all cracking in che 10-inch schedule 40 type 30't inless dischar"e pip=ng of the "A" recirculation spray beat exchanger a" Su~ L<"'t No.
2.
A recent in"paction of Un' Containnent Reef.rculatian Sp y
Pipi..g revealed cracking similar to Unit 2.
Oa October 8, 1976, -no"her incide"t af similar cracking in 8-in"h schedule 10 t5we 304 stainless piping of the S."faty Xn5ection Pt.np Suction Line..t the Ginna facility wee reported by the licensee.
Xnfomat~n received an th" netallurgical ana1ys s conducted ta d-te indicates that the failures vere the result 'of i-.cergranuler =t"ess corrosion crac~F. th-t initiated on the inside of the piping.
A comonelity of ctors obsewed as"ac~ared with the corrcsio necha=i>
were e 1.
The cracks;-are ad-'acent to and p opag ted along veld zones of the thin-walled 2.ov. pressu e piping, not part of the reactor coolant system 2.
Cracl&g occurred in piping containing relatively stagnant.
or='
acid so 'tion not equ" red for no~Gal operating co ldll t~onse 3.
Analysis of surface p oducts at this tMe "ndicate a chloride ion inte action with oxide formation in the relatively stagnant bo Xc acid solu=ion -s the probable carrodant,
~Cth the state of s"ress probably due to velding and/ar fabrication.
The source of he chla=i<e ion is nat definitely icnov>.
Saw@ver, AHO-1 the ch".orides and sulfide level observed in the su face taw~sh
<<~~ nea" velds is beD~eved ta have been introduced inta the p ping during testing of the sodium thiosulfate discharge valves, o
valve leakage.
S~a ly, at Gin-a the chlorides and potenti-T o~g"n
XE Circular Hp. 7&46 Nove=ber 26, l976 availability were assumed to have been present ince original constmction of the borated vater storage tank which i-vented to atmosphere.
Corrosion attac~z at Surry is attributed to in<<3.cadge of chlorides through recirculation spray heat e>:change tubing, allo>>~ng bu~Mdup of contaminated water i" an other ~se normallv drF spray n$ n$ ns-hCTXOÃ TO 3r. M~ BY LXCZRSFF.:
Provide a description of your program fo" assuring continued
~ integ ity of tko e safety-related piping systems which are not frequently flushed, or which contain nonflo.Cn" liquids.
This program should include consideration of hydrostatic t ting in accordance Mth AB:-K Code Section XI rules (1974 Hition) Eo all active system required for safety injection and containment spray, includ'"g their recirculation modes, from source of water supply up to the second isolation valve of the primary s>%tern.
5~~3ar test
- should, be considered for ocher safety-relat d pipin syst~.
2.
Your program should also consider volumetric examination of a representative number of circus; creat" a'ipe welds by non-destructive eraa~na.tion techn-ques.
Such e"a-"nations sho ld be'performed generally in accordance with hppendSx l of Section XX of the AS'E Code, evcrpt that the examined area should cover a distance of appzov~ately sir (6) t~~s the p"pe val3. thickness (but not less th n 2 inches and need not exceed S inches) on each side of the weld'upplementa~
eza-~mation techniques, such as radiography, should be used where necessary fox evaluation. or confir=ation of ultx sonic indications resulting fron such exatnination.
3.
A report describing your program and schedule for these inspec-t-'ons should be submitted within 30 days a-ter receipt of this Cireu2.ar.
4.
The NRC Regional 0 fice should be infoaaed within'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, of any adverse findings resulting during nondestructive evaluation of the accessible pip~~p welds identified above.
5.
A, su~ay report of the examinations and evaluation of result should be submitted within 60 day" from the date of co~letion of proposed testing and examinations.
L~ Circular Joi 76-06 Hovember 26, 1976 This s~ry repor ho&~ d also include a brief description of plant conditions, operating procedures or other activities vhKch provide ns"urance thar. the eHluent chemistry vQl maint in.
3.ov levels of potent&il corrodants in such relatively "" gnant regiors within the piping.
Your responses should. be submitted to the Director of this office,
-ith a copy to the HRC Office of Enspection and Enforcement, Division o
Reactor Iaspection
- Programs, Lia hington, D.C.
20555.
Approval of iQC requirements fox reports concerning, possible generic problems has been obtained under 44 U.S.C 3152 from tne U S-General Account~kg Office.
(GAO Approval 8-180255 l',80062),
axpires 7/31!77. )
I
~
'IF. Bulletin Ho. 79-17, Revision 1
October 29, 1979 LicTIli6 OF IE BULLETINS ISSLiED if/ LAST SIX MONTHS Enclosure Paoe 1 of 3 Bulletin No, 79-24 79-23 Subject Frozen Lines Pot'ential Failure of Emergency Diesel Generator Field Exciter Transformer Date Issued 9/27/79 9/12/79 Issued Tn All power reactor facilities which have either OLs or CPs and are in the late stage of construction All Power Reactor Facilities with an Operating License or a construction permit 79-14 Seismic Analyses'or (Supplement
- 2) As-Built Safety-Related Piping Systems 9/7/79 All Power Reactor Facilities with an OL or a CP 79-22 Possible Leakage of Tub s 9/5/79 of Tritium Gas in Time-pieces for Luminosity To Each Licensee who Receives Tubes of Tritium Gas Used in Timepieces for Luminosity'9-13 (Rev
~ 1)
Cracking in Feedwater System Piping 79-02 Pipe Support Base Plate (Rev.
1)
Designs Using Concrete (Supplement
- 1) Expansion Anchor Bolts 79-14 Seismic Analyses For (Supplement) 's-Built Safety-Pelated Piping Systems 8/30/79 8/20/79 8/15/79 All Designated Applicants for OLs All oower Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL or a
CP 79-21 79-20 79-19 Temperature Effects on Level Measurements Packaging Low-Level Radioactive Haste for Transport and Burial Packaging Low-Level Radioactive Haste for T>r~<soort and Burial 8/13/79 8/10/79 8/10/79 All PNRs with an operating license All Haterials Licensees who did not receive Bulletin flo. 79-19 All Power and Research Reactors with ALs, fuel facilities except cc'r 2 >'.,p+
1 i c~<> reps
IE Bulletin No. 79-17, Revision 1
October 29, 1979 Enclosure Page 2 of 3 Bulletin No.
Subject LISTING OF IE BULLETINS ISSUED IN LAST SIX fMONTHS Date Issued Issued To 79-18 Audibility Problems Encountered on Evacuation 8/7/79 All Power Reactor Facilities with an Operating License 79-05C806C Nuclear Incident at Thr e 7/26/79
!iile Island - Supplement To all
~llR Power Reactor Facilities with an OL 79-17 Pipe Cracks in Stagnant Borated '1later Systems at Pgp Plants 7/26/79 All PMR's with operating license 79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Power Reactor Operating Licenses who anticipate loading fuel prior to 1981 79 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 Piping System Cracking In Feedwater System Piping Pipe Support Base Plate Designs Using Concrete Expansion 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 Pl/Rs with an OL for action. All BMRs with a CP for.
information.
All Power Reactor Facilities with an OL or a
9 I E Bulletin Ho. 79-17, Revs ion 1
Octoher 29,- 1979 LISTIHCi OF IE BULLFTIHS ISSUED IH LAST S IX hOHTHS Enclosure Page 3 of 3 Bulletin No.
Subject Date Issued Issued To 79-01A 79-12 79-11 79-10 Environmental qualification of Class 1F. Equipment (Deficiencies in the Envi-ronmental qualification of ASCO Solenoid Valves)
Short Period Scrams at O'AR Facilities Faulty Overcurrent Trip Device in Circuit Breakers for Engineered Safety Systems Reoua1 ificati on Trai ning Program Stati sties 6/6/79 5/31/79 5/22/79 5/11/79 All Power Reactor Facilities with an OL or CP All GE BMR Facilities with an OL All Power Reactor Facilities with an OL or a CP
~
All Power Reactor Facilities with an OL
s~