ML20086E500
| ML20086E500 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 06/13/1974 |
| From: | Riley B VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20086E492 | List: |
| References | |
| AO-74-08, AO-74-8, VYV-3303, NUDOCS 8312070342 | |
| Download: ML20086E500 (13) | |
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P. O DOX 157 V c H N OtJ. VERMONT o33%1 June 13, 1974 Director Directorate of Re:,ulat ,ry Operations, Region I -
United States Atemic I .rgy Commission 631 Park Avenue King of Prussia, P. ;ylvc la 19406 ,
REFEnr5CE 0,crat: q License D .(-28 Docket . 50-271 Abnomal r currence !!n. A0-74-08 (Spec!al 20 Day Report)
Dear Sir:
As defined by Tcchnical Specifications for the Vermont Yankee h'uclear Powe r S t. t ic , S ctir, 6.7, we are reporting the following Abnorral Occurrence a:, 'AO-74 3.
At l '40 on !!ay 25, 1974, with the reactor operating at approximately 10% pouer plant personnel who were performing corrective maintenance on g
reactor dry 1 11 cooling equipment observed pater in some ventilation ducting and at soue instrument penetrations at the biological shield.
Initial investigation indicated that the source of leakage appeared to be f ron within the biolegical shield rea. An inmediate reactor shutdown and cooldern uns initiated in order to define the source of the leakage.
On ! ay 26,1974, a uater leak was identified on t he Control Rod Drive 11ydraulic Systen return line to the reactor vessel. The leak was located on a 2 1/2" x 3" stainless steel butt .;cid concentric reducer which connects the systen piping to I-9 ve::sel no: zle " safe end" and uas initially determined to be in the area of the fusion line of the reducer sidc of the veld. Figur 1 cont. Ins a schenatic representation of the versel and piping in the affected area.
It was concluded that this condition was report.nle under Technice! Specifications, & ction 1.0.A.5 in that it resulted in abnornal denrade. tion of one of the several boundaries desi",ned to cont a 'n radioact ive natorials.
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4 Direct orat e of Ee ulatary Operat ican June 13, 197'.
Page 2 An immediate progran as initiated by the Yankee Atonic Electric Company Nuclear Service: Division (USD) which included an in-depth engineeri: revicw of the cause oft 4 problen and the necessary requirements to perforn exacting coi ' active action. Thi:, report vill sunmarize the ie igati,n and subsequent resulting corrective measures.
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N The defective redt ..er including the weld at each end was cut out of the line. The reducer, includin g the defect and the associated end welds, were sent to Battelle Memorial Institute, Col mbus L." . iric , for sect ioning and analyr -; . The following discussion of crack norphology is basc ' on the findings by Eatte'le.
The analysis it 'cated that the crack had originated at a significant s eens r aer on J in ide diamete. of the reducer at the larger end adjacent to the re ucer to safe ad weld. The stress r!ser ,
cont 'sted of a nachined not ch with essentially ne radius of curvature nd a slope greatly in excess of the code required taper.
Figure No. 2 cc a': a schematic representation of the notch including act u:. dimensions. Figure No. 3 provi ,as an example of a'codt acceptable decign. Inspection of the two figures readily _
reveals t 3 signif:cance of the . notch in locally concentrating the stresses. The crach originated at the notch and propagated circunf(rentially around the pipe following the notch while
'proparating radially outward eventually penetrating the outside -
dinneter of the reducer. The length of the crack at the outside wall was appro::irnt ely 3/4" centered at the 12 o' clock position. h The crack length at the inside wall was oi, the order of 2 1/2" and extended from approximately the eleven o' clock position to the two o' clock position when viewed fron pipe to vessel. Initial evaluation of the crach surface indicated that fatigue was a prinary -
failure ncchaninn; houever, the f racture surface wa: partially obscured by a tight ly adherent oxide filn. The fratture surfaces were descaled and .. examined. The subscquent reexamination revealed that r.ther than fatigue, th .f prinary nechanism for failure 9
appeared to he(transgranular stress assisted corrosion crachi n This conclusion EE~7eiiched prina-fTy because of the quasi-cleavage appearance of t he f ractu re surtac . The internit tent st riations ,
which origina11 were thought to be fatigue oriented were actually deternined to .epresen. Landed struct ure of the corrosion product.
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June 13, 1974 I
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There were at le: t ten of these unevenly spaced baads of heavy corrosion ptodue; build-up which verified the intermittant nature of the failure. TL e ridges uere basically randou in orientation although sonv atual fatigue striations appeared toward t' e edge of the crack tip in the latter stages of crack growth as vould be expected. There was also evidence of cold work in th vicinity of thc net ch which would be expected f rom the machining A e in that arc .
There was evidence of secondary crack grouth i :.ernittently running off the main cratk; there cracks e re also transgranular in naturt. During the final stage of the f.ilure, the main crack did protrude into the heat affected zone of the reducer and ultinately to the veld itself; however, weld-crack interaction is not considered a factor in the failure.
The reducer bane naterfal is not sensitized except for some nornal veld ser'itinilon in the heat affected zone. The salient point of this .' ' mis is ' hat the cr.ek originated within the notch root which ; outside th..
heat affected zone of the reducer.
The corro: "n pr( ?.uc t r present in the crack were examined for chemica] t.astituent.. Results indicated that chere war no chloride or caustic present. The oxide film consisted primarily of the two iron oxides, re2 03 and rc3 04 The phenomenon of transgranular stress assisted corr.sion cracking at a stress concentrat on point like a notch in austenitic stainless steci in the presence of water and oxygen has been discussed in several papers. The TMR environ ant in the pip:ng system under consideration may contain oxygen IcVels in the vic nity of 0.3 - 1.5 ppm. The extremely high stresses prese: at the point of failure will be presented in a later section in this report. The phenomenon of stress assisted corrosion cracking is discussed and documented in the following documents:
- 1. "Intergrruular Stress Assisted Corrosion Cracking of Austenitic Alloys in h'ater Cooled Nuclear Reactors,"
by Craig F. Cheng - Argon National Laboratory, June, 19/3. Pres.nted at the fif t h European Congress on Corrosion in Paris, France on September 24-28, 1973.
k'ork'vas performed under the auspices of the U.S. Atomic Energy Commiusion.
- 2. " Stress Corronic Failurt ," by Peter R. Suann -
Scici;ific American, February 1966.
II. REPA T H SEnU!'.':CE The repali pi ngram was carried out in accordance Ith the 1974 edition of the ASME nuller and Pre? ':ure Vessel Code,Section XI; nule. for In-Service luspect lon f Nuclean Power Plant A._uum -
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Dir.etorato of henul nry Operat f on-June 13, I l'4 Pane. 4.
The repair included replacerent of the existing reducer with a new certified safety clans o. reducer of equivalent naterial. As part of the repair effert, the heat affected tone on the stainless steel safe end and tbc 2 1/2" piping section downstream of the original reducer'were machined out. The re f or.' , in order to maintain the same overall pipe length, a 3" diameter, 3" long spool piece was inserted between the safe end and the new reducer. The 2 1/2" pipe was cut back by the amount necessary to re-est ablish the original pipe configurat ion length. Figure 4 shows' t he original pipi' ; geomet t, in the area of the ceducer. Figure 5 shows the new inst allation as discussed above. The new reducer does not have gr.. nach: ed on tla ID ai ' consequently there are na stress concentra: fat its associ: ted with the new installation which are not r ndy accou ted fo. in the original stress calculations.
All weldi , dr.ne durin, the repair was accomplished ie accordance with properly qualific. procedurc~ as required by Section XI.
Non-destructive "xamination included liquid penetrant testing of the root arca av final sut .' ace of the new weldments and radiography of the final veldmente. These non-destructive c::aminations were also condu in accordance with properly quali fied procedures as requir( d by Sectici XI. Subsequent to the completion of the repair, a hydro.ratic test u s performed at a pressure of 1.02 times the nominal perating pre: m e(which corresponds with 100 percent rated reactor power) at a tempt ature of 500*F. This hydrostatic pressure is in accordance with the requiren its of Section XI, 1974 edition.
The test press re was maintained for a minimum of four hours prior to performance of tl..: visua examinations. The examination included specific visual checks of the new welds cnd a general inspection of all other drywell piping. Subsequent to the successfal completion of the hydrostat.ic tests, a neu ultrasonic base line examination was conclucted on the new welds in accordance with the procedures of Section XI.
In addition, an inspection agency (llartfo.1 Steam Boiler Insurance Company) was ceitracted to review the overall repair program and procedures utilir.e. as required by Section XI.
As part of the repai r e f fort , the piping rest raints on the af fect ed line vere c' < ked and dete nined to be properly in. talled in accord. c w.th the Ehasco pipe flexibility analysis except an nott.1 helow. Motion indicat ed by abrasion narks on the rest raints ind'ented that Q D llydraulic Return 1.ine motion was consistect wit h t hat predicted by the Ebance pipe flexibility analysis.
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l Directorate ofth..1 ry 0;.crat ions June 13, 197!.
Page T During th e xa-ina t i on , it was determined that an additio.a1 lateral restraict existed an shown in Figure 1, rest r 'ot 1.
This installatica had no effect. on the stress level; paesent in the line reculting fron vessel vertical growth, since it did not restrict notion in that direction. Nevertheless, because it was not part of the <.riginal design, it was renoved. The other restraints and hr..q,crs in the line were determined to be properly installed and are indicated in Figure 1. .
Ill. STRESS II. .E1.S PRESE':T AS A RESUI T OF MOTC' This sectien of thi repr ~t c. atain's the re: :ts of an analysis of the state of stress r ting in the reducer to safe end weldment
'and re'icer notch are prior to failure. In order to provide an evaluatien of the ac' n.1 stress 1c .cis pr. nt, hand calculations were con?ncted by the Euclear Services Divisi in accorda;'e with the tmt' Os and 'rocedures of the AS:!E Boiler ..qd Pressure Vessel Code. Se. ion III; Nuclen. i oc r Plant Conponents, 1971 edition.
Stresse utre calculated for t12 heatup, cooldown, and 1teady state coditions including stressen due to pipe loads and internal pressure. In arder to wess the affects of thernal gradients in the regic of the reduc.:r, seven thernocouples were installed on the piping prior o startup, as shown in Figure 5. Temperatures
'were nonitored during the post repair startup trannient and steady - .te operat:on. The gradients which existed across the region repre atative of the original reducer location were utilized in performing an interaction analysis of the safe-end-reducer intersection in accordance with the nethods and procedures of ASME Boiler and Pr ssure Vessel Code Section III, 1971, Article A-7000.
The re 'ltan' thernal gradient stressen are included in the stress valucc p estated in sections A and B helow. The vessel nozzle stren es included in ti calculationn were taken from the Vernant Yankee. Reactor Prcm:ure Vessel _ Stress Report 9-6201-1, Revinion 1, dated 6-10-69. Piping forces and nonents were taken from the Ebasco CRL . Piping Flexibility Analysis. The data presented below in provided in two categorien. The first shown th. effect of the notch in raising the strens levelu, the second portion shows the stress levels present in the line without a notch and are representative of values pre:ent in the repaired section of pipe. It can readily be n 'n that te ef.ect of the notch (Section A) is to raine the stri : Jevils to permit onset and propagation of stress assisted corronien ' aching as dencribe i'n Section I of this report. It is readily arent from t'" necond set of numbers (Sect ion 11) that
. the original .enign and the prenent ecptivale nt confinu':.t ion have w Auva . _ _ _ _ _
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Directorate of negulat cry Operat i: ns June.L3, 1974 Page 6 A. Mtch Effects
.The strer concent etloi. factor associated with the notch shown in ri; ue 2 was calculated to be 4.74.
This factor was d termined in acco: 1- .cc with the t ethods and procedu.es of Welding Research Council Bulletin 107, 1968, Appena b: B, 1.ocal Sttacscs in Spherical and Cylind ric:- : lhe11< due to Loads, by Wicht:an, Metshon, and
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Hopper. .ie dir c ; ions utilized in the calculations were taken directly from the failed specitNn.
The peak sti s intensity in the vicinity of the notch was calculated to be 154,000 psi. This is considered primarily a sustait ed load, such that the notch area would be subject d to thi: value contino sly during service. The naterial proper '. i. of the reducer, which is Type No. 304 stainless steel are as follows: a ninimun 500*r yield strength value of 19,440 psi. aad an ultit-ate teasile strength of 57,750 psi.
B. Linconc :tratnd (no notch) St ress Levels Present in the Orini al Dej.n and Present Pip [ m Confinuration of CRD Hydraulic Return Line
. The pcak ..ress intensity with no notch present is calculated to be 42,460 psi which is within the code allowable limit of 52,200 psi, based on t:aterial properties at operating temperatures.
The above calculationr, were essentially hand calcul:* ions and no forme.1 computer programs were utilized. They are, nevert' . lens, considerci' to be accurate within the limitations of hand calculations and adequau to properly assess the cause of failure and tle repair. As discussed in Section V, an indepet.!cnt stress analysis utilizing finite ela.ent computer code techniques has been undertaken. l IV. IMMr.DIATE C0" g RIVE PROGRAM l
The program: discussed in this and the following section are the combined rer.n1tr. of the reco:x ?u. Lions of the following groups, which revis .ed the rt .uir prot, ram and cor ective prc,cedure::.
A. Yanhec 1: .:len Services Di-Isloa, Mechanical Ent,ineerin:'
Group
h . ..' ,_ 4 VERf.10f "r YAtJNCt NUCLCAN t' .Wtn Cour op .- -
Directorate or Regulatory Operations June 13, 1974 Page.7.
The follouing step were taken to insure thit there was no unresolved safety question prior to plant return to power operation.
- 1. A pi..e r'straint inspection was conducted by Ebasco of the following syste in the drywell: Core Spray Syster, RHil, Reactor Ci . anup , liPC1, Feedwater , RCIC ,
Main Stean Syster. Th inspection program cot: pared the existing rest raint installation to that s;ecified by the Ebesco st. v. isoiatric drawings 2.nd checked to ensure that restra'nts were properly located, tint therc were no additional restraints, and that there were no temporary supports. The inspection revealed no discrepancies. In addition, Ebasco checked pipe hangars for proper i 4tallation on the CRD hydraulic return line, core sp.ay line, and ot' er sy
- cms rand." ly selected.
There ucre no discrepanc~,s forad execpt for the one lateral ristraint on the Cia line as p*cviously discussed.
- 2. The manufacture of the reducer was detert..Ined to be
. Flow-Line, New Castic, Pecosylvania. The spare parts system at Vermont Yankce was checked and two Flow Line stainless steel ' 1/2" 90* elbows were found in stock.
These were re: aved ,d checked for the presence of a notch
- similar to that of the failed reducec. The inspection revealed tl presence of what appeared to be a properly machiw d counterbore in a location similar to that on the failed reducer. In order to verify that the bevel va.
properly machined to code requirements, one of the elbows war. sectioned through the bevel at several locations arcur.d the circumference and measurerm nts were taken of the bevel and root radius. The re its indicated that the bevel was made at appro::ir ately 14 angle with a root radius of
. 050 inch. These values are consistant with code allcwable designs and are typical of the weld preparation bevel shown in Figure 3. These valres for bevel and root radius stand out in rarhed contrast i those values for the failed reducer; specifically, the failed reducer exhibited a notch of 80* wit h a root radius of .0014 inch.
- 3. During the by ostatic test, all drywell piping was examined for evidence of leakage and none was found.
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VCI'4MONT YAf JKCC NUCLCAH POV C0f u'O P/. T Directorat of Reguledory Operations .
Jun 13, 1974 page 8
- 4. Drywell piping r*tached to the reactor vessel was checked to detei ,ine whc her there were any similar piping configtn ations. The inspection revealed that there were no similarly config: d reducer to none'e welds and no piping configuratloas which would be subject to reactor
- v. ! vcirtical growth end the resultant stresses.
V. LO"C RY .J_. _.P. J 0 G RN f
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As dircussed in Section IV, the program outlined herc includes the recon:
ndations of the groups and committees previously listed.
- 1. Teledyne P ~:crials Researcli of Waltha'a, Mr. sachusetts has been contr :ted to conduct an indepet ent stress analysis which will. includ a finite cler nt toteling of the specific notchc region unde, invc 'gation. T..e basis and the criteraa of th::t analynis will be consiste..L with the Boiler and Pressure Vessel Code, Section Ill. It is anticipated that this re, art will be available within approximately six week .
- 2. The follo.ing suppletentry work will be performed at the next schedaled reft ling, outage (presently schedu]cd for October, 1974).
. a. A detailed inspection program will be conducted to verify the integrity and proper installation of all pipe angars and restraints within the Primary Containment.
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- b. An augiaented i service inspection program which will Inc.lude surface and volumetric examination will be conducted i the following areas:
, 1) All reactor vess 1 s,fe end to nozzle welds.
- 2) All vessel s: end to pipe welds.
- 3) A minimum of tlnce additional welds in all horizontal pipe rc, originating at a reactor versel noz..le, between the vessel and the first elbow in the pipe cun.
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- June 13, 1974 Pa ge . 9-VI. SU'PL\nY AND COSCLU310'M l Based en our inzestigations and the information presented in I
this repo: r , it is our conclus. n that the failure was caused by the presence of a notch whica~ raised the stress levels piesent to such a value that stress associated corrosion cracking became p o sr. : ble . Discussion of this type of phenomena are contained in papers referenced in an earlier section of this rep 7rt. The notch acted effectively as a crack st art er and was able to propagate as a result of the concentrated stresses it, a suitable environment.
Without the notch present, it is our position that the stress 1evels pres.It are well within code allowable values, and that therefore, no failures of the type which occurred are possibic in L12 repai rs .' section of pipe. There is considered to be no unresolv,d saft. question in that the refair does not create the pot <atial fo:- a new vecident, the probability of occurrence of the same acci :nt is not increased since the repair resto es the piping to a configuration which it the equivalent of the oririnal wi no nqtchlpreseqt, and the margin of safety as d r.;ign at e v in tl. basis of any technical specification is not reduced. When the Tcledyne study is completed, ue will at' ise you of those rer.ults. Record:, derived from the additional long range progran .11 be avstlable'on site after the next scheduled refueling outage.
Should you desire any additional information, please do not hesitate to contact us.
Very truly yours, VER:10NT Y. 'KEE NUCLEAR POWER CORPORATION
- s. , .( h ad h B.W. ti .l ey Plant Su, rintendent TDK/RUS/kbd
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