BVY-92-055, Advises That Util Performed Rv Head Insp During Outage to Address Concerns Raised in GE SIL 539 & Discovered Possible Cracking in Cladding of RPV Head.No Evidence That Structural Integrity of RPV or Vessel Head Compromised.Evaluation Encl
| ML20091D380 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 04/05/1992 |
| From: | Murphy W VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-92-055, BVY-92-55, NUDOCS 9204090344 | |
| Download: ML20091D380 (12) | |
Text
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i ERMONT YANKEE
' NUCLEAR POWER CORPORATION
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Ferry Road Brat %oro, VT 05301-7002 Warren P Murphy gjt sy I Senior Vice President. Operations
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(802) 257-5271
\ BVY 92-055 April 5,1992 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
Reference:
a) License No. DPR-28 (Docket No. 50-271)
De'r Sir:
Subject:
Proposed Alternative for Compliance with 10CFR50.55a Regarding RPV Cladding Indications During the current 1992 refueling outage, Vermont Yankee performed a reactor vessel head inspection to address concerns raised in General Electric SIL No. 539. This SIL discussed the situation at Quad Cities Unit 2, where cracking was detected in both the stainless steel cladding and the low alloy steel under the cladding of the reactor pressure vessel head. Vermont Yankee discovered indications in areas of cladding inside the reactor vessel head. Additionally, during the 1992 refueling outage, Vermont Yau ee utilized a new technique (remote color video inspection camera) to perform visual inspection of reactor pressure vessel (RPV) internals. This visual inspection revealed similar cladding indications, primarily in the shell Gange region of the vessel.
The purpose of the stainless steel reactor vessel cladding at Vermont Yankee is to reduce the presence of iron oxide products in reactor coolant, thus reducing the demand on reactor coo' ant demineralizer equipment to maintain water purity. Vessel cladding does not form a part of the reactor pressure vessel structure and no credit is taken for it in the structural analysis of the vessel or vessel head.
Vermont Yankee has performed extensive ultrasonic examination techniques of over 140 indications and employed manual exploration of a typical " worst case" visual indication to investigate the nature of the indbations. As a result of these investigations, we have concluded that the indications are the result of stress corrosion cracking of the cladding, and there is no evidence that the indications compromise the structural integrity of the reactor pressure vessel or vessel head. This conclusion has been confirmed by the volumetric and other examinations described in the enclosed " Engineering Evaluation".
The inspection of the RPV head had been performed in response to GE SIL 539. If the inspection was being performed as part of the inservice inspection program of Section XI of the ASME Code, the acceptance criteria would he obtained from IWB-3520. Since the 1980 Edition of the Code, which is the applicable Code Edition for the Vermoat Yankee ISI program, does not provide any standards for the evaluation, the 1989 Edition of the Code was consulted for guida.nce. Paragraph IWB-3520.2 provides ;
acceptance standards in the 1989 Code.
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, ; U.S. Nuclear Regulatory Commission ;
. April 5,1992 - VERMONT YANKEE NUCLEAR POWER CORPOR ATION Page 2 of 2 i
l IWB-3520.2 requires that relevant conditions be corrected in accordance with the requirements j of IWB-3142. Paragraph IWB-3142.2 allows acceptance of relevant conditi-is by supplemental ;
examination, and directs the user to IWB-3200. Paragraph IWB 3200(b) states, " Visual examinations that ;
reveal relevant conditions described in the standards of this Article may be supplemented by surface or
- volumetric examinations to determine the extent of the unacceptable conditions and the need for corrective f measures, repairs, analytical evaluation, or replacement " ;
i As long as it can be determined that the extent of the supplemental examination is adequate to provide assurance that there is no need for " corrective measures, repwrs, analytical evaluation, or
- replacement," it does not appear that the Code requires an evaluation of 100% of the indications.
Ultrasonic inspections (UT) have been performed in all representative areas of the reactor head l and in the reactor vessel near the head flange area. In excess of 140 indications were interrogated by UT, with no indication of basemetal penetration. Thirty-two of these indications were in the lower vessel flange and upper shell course area. The inspections performed by Vermont Yankee provide a high level of confidence that the regions inspected reflect all areas where similar indications exist in the re etor pressure vessel cladding.
This condition was discussed with NRC NRR and Region I of0cials on April 1,1992 at which
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time the question of an alternatives application under 10CFR50.55a(a)(3) was considered. This !
application is submitted in response to a NRC suggestion due to the uncertainty regarding the application of the ASME Section XI Code to this situation. Therefore, this letter formally requests NRC approval of a proposed alternative to Section XI of the ASME Code.
The specific alternative being proposed is to allow sampling of conditions believed to be benign rather than requiring examination of all indications. The sampling performed by Vermont Yankee
- provides a high level of con 0dence that the cladLg indications do not penetrate into the basemetal.
The attached report provides a detailed discussion documenting the technical acceptability of the i proposed alternative.
We request your prompt approval of this request. Should you have any questions regarding the attached information or require further information, please do not hesitate to contact me directly In
- addition, should you Hnd it necessary, we are prepared to meet with the NRC staff to present our Hndings on short notice.
Very truly yours, Vermont Yankee Nuclear Power Corporation j-
- 1
, R.
Wj Murpi y [
Senior Vice Pres kng perations I cc: USNRC Region 1 Administrator USNRC Resident inspector VYNPS USNRC Project Manager, VYNPS m
ENGINEERING EVALUATION OF VERMONT YANKEE REACTOR PRESSURE VESSEL CLAD INDICATIONS 1992 REFUELING OUTAGE
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I likCKGROUND l
in November 1991 GE issued SIL 539, which recommended that plants with stainless steel clad reactor pressure vessel heads inspect the area corresponding to the Quad Cities cracking. This SIL discussed the Quad Cities plant, where cracking was detected m both the stainless steel cladding and in the low alloy steel under the cladding of the reactor pressure vessel head.
A visual inspection of the internal surfaces of the Vermont Yankee reactor pressure vessel head was performed on March 17,1992 in accordance with the secommendations of the GE SIL No. 539. Numerous areas of rust on the stainless steel cladding were noted. This would be indicative of clad penutration down to the alloy steel basemetal. The rust areas were seen at the area of the flange to head butt weld and on the manual cladding on the flange forging [see Figure 1). Additional rusting was t.bserved on the manual cladding of the dollar plate section of the head. These areas were mostly limited to locations of manually applied cladding; however, some rusting was also visible in the area of automatically applied cladding. Some of the areas of rust were associated with visuallinear indications. The majority of the rust marks are onihe flange area with a long indication at the manual-to-automatic clad weld interface. Approximately 2/3 of the circumference at that interface shows some rust. There are also some indications on the manually clad area of the dollar plate. This area is about five feet in diameter at the top of the head. Most of the rust marks follow the circumferential clad weld heads.
In addition, during visual examination of the RpV internals, similar rust areas were observed on the vessel cladding. This type of indication was visible due to the use of a new remote color video inspection camera. The vessel cladding indications were visually similar to the head cladding indication and were primarily in the vessel llange region of the vessel.
Vermont Yankee immediately reviewed the conditions seen at the Quad Cities 2 power plant for any similarity to Vermont Yankee. Discussions were held with Commonwealth Edison, General Electric Company and Structural Integrity Associates (SIA). SIA had performed evaluations of the Quad Cities RPV head for Commonwealth Edison and EPRI, These discussions revealed that the Quad Cities RPV head was experiencing two different conditions. The cladding developed intergranular stress corrosion cracking (IGSCC) attributed to lov. ferrite content, welding residual stresses, cold work stresses from grinding and an aggressive environment.
The Quad Cities RPV head basemetal developed cracks from a mechanism known as reheat cracking. This is caused by reduced material ductility resulting from alloy impurities along with residual stresses from coldwork and differential thermal expansion. The high heat input strip clad process used h't Babcock and Wilcox (the fabricator of the Quad Cities RPV) in the early 1970's is now known for producing reheat cracking in certain low alloy steels.
The metallurgical samples taken from Quad Cities showed evidence of stress corrosion . racking in the clad and reheat cracking in the basemetal. In some cases the cracks meet at the clad /basemetal fusion line. It is important to stress that there is no evidence at Quad Cities that the IGSCC cracks progressed into the basemetal.
VERMONT YANKEE CONDITIO_N The following was observed and concluded regaruing the conditions on the Vermont Yankee reactor pressure vessel head:
fabrication Procew None of the cladding on the Vermont Yankee vessel was applied by the high heat input strip cladding process used on the Quad Cities vessel. The automatic clad welding process used at Vermont Yankee required only one layer of clad. The manual clad welding process re'luired two layeis; more could be applied if required to meet the nominal clad thickness requirement of 3/16 inches. During the welding process the basemetal layer 1
l melts and mixes with the weld metal that is being deposited. As the welding heat input is increased, a greater
! amount of basemetal is melted. This results in an altered chemistry of the weld deposit, as compared with the original welding electrode. Depending on the heat input, this rolloying proecss can occur in mu!tiple layers of weld deposit. Studies performed in 1986 during the development of the core spray nonle weld overlay at Vermont Yankee showed that a minimum of three weld layers were needed to obtain an unaffected weld material chemistry sample, and this was from a very tightly controlled heat input welding process. Similar experiences are reporoA.! in the technicai literature. An engineering mvestigation performed as part of the Vermont Yankee reeirculation system pipe replacement project in 1985 showed that as carbon content increased above about 0.05 w/o the amount of ferrite required for IGSCC resistance increased rapidly [see Figure 21. Therefore, it is likely that the cladding has low IGSCC resistance.
The post weld heat treatment [PWHTl process for the Vermont Yankee vessel was evaluated to determine if the fabrication prt.eess could have reduced the ferrite conte ' of the RPV head claddirg. No significant effect on the feriite content would be expected.
The RPV head tlange is a SA508 CL2 forging. The head plate material is SA533 GrB. Structural Integrity Associate 3 reviewed the material certifications for the head tlange and plate material and concluded the forging could be susceptible to reheat cracking. llow ever, reheat cracking would only occur if significant welding residual stresses, cold work or differential thermal expansion stresses were present. The fabrication records do .
not indicate this to be true, therefore reheat eraeking is unlikely. Extensive ultrasonie examination of the head and shell llange regions, as described below, demonstrate that :here is no cracking in the basemetal.
11SPECTION RESUI.TS The investigation techniques s 3d at Vermont Yankee, visual, penetrant, and ultrasonie, followed the (nommendations of the SIL. It is understood from discussions with Commonwealth Edison that the ultrasonic tedniques were revRwed by the NRC for Quad Cities and were qualified at the EPRI NDE Center by petformance demonstration on controlled tlaw samples.
. Penetrant Tt dtng imd Manual Exnloration .
Two one foot square areas on the head flange were penetrant tested. The indications appeared to closely resemble the size, frequency and pattern of the rust indications. An investigation to determine the depth -
of an indication was then undertaken. A typical " worst case" (from visual appearance) indication was selected. A length of about 3 inches of the indication was excavated. At steps of 1/16 inch or less in depth, a penetrant test was performed to determine if the indication was still present. As the cladding was removed, and the depth increased, the indication appeared to break up into more numerous tiner indicatbns and grew tighter and more diffuse, which would be indicative of IGSCC. At a depth of almost 3/16 inch. the indication had almost disappeared, except for some very faint areas, which upon closer inspection appeared to be tight craze cracking. These faint indications remained within the cladding,
=
Ullrasonic Examination UT evaluation of a number of indictilons, including the "ne that was manually explored, contirmed that the flaw depths were contained witt in the clad thickness.
The clad is nominally .19 inch thick, and is specified to be .125 inch minimum with no maximum specitled. An original piece of vendor supplied vessel material with cladding (the RPV vessel ultrasonic calibration block) is available. The clad penetration on this block measures between 0.125 inch and 0.375 inch thick. This was measured directly off of the block cross section anJ also with a high resolution ultrasonie straight beam search unit.
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Ultrasonic inspections were performed in accordance with the GE SIL recommended techniques. These techniques successfully detected and sized the 2% notch on the RPV vessel calibration block. Various locations on the inside surface of the head flange, the dollar plate, the vessel 11ange, and the upper shell course were examined. UT was also performed from the OD of the head. in excess of 147 redectors were measured by UT, with no indication of basemetal penetration. The depths ranged from 0.08" to 0.30" Several of the deeper reDectors were re-evaluated with the straight beam search unit to determine the clad depth at those locations, and con 0rmed the cladding was thicker at these points.
During the visual examination of the RPV internals, a rust area was observed adjacent to the fillet weld where the steam dryer support bracket weld meets the vessel cladding. This indication was quite visible since the in-vessel examinations were being conducted with a color camera. The 1983 black and white video tape of the same area was reviewed, in the same area as the rust, one could discern a slightly lighter region. This was not recorded as a region of concern in 1983 and if not for the increased contrast of the color view in the 1992 inspection, it would be hardly noticeable. The area was thoroughly UT examined from outside the vessel. No evidence of any flaw propagation into the basemetal was detected (see Figure 3].
= Ferrite Testing Ferrite readings were taken at 24 locations on the head Gange inside surface and 12 locations on the vessel Oange insido surface. The lowest reading was 5% to 7.5%, with the majority of readings at 7.5%
- a 10%. This provides evidence that the cracking in both the RPV head clad and the vessel llange clad is due to IGSCC and not micro-Ossuring which can occur in low ferrite, in the 0-3% range, stainless steel weld metal.
- Dhemieul Analysia Metal chips were removed from the head and vessel llange area cladding for chemical analysis. The carbon content of the material removed was 0.079 w/o for the head dange and 0.11 w/o for the shell 11ange. The carbon contert of the shell Hange cladding is well alvave the specification upper limit for carbon of 0.08 w/o max. The head ihnge vable is within specifier. tion range. However, a review of all the welding electrode material certineations showed no heats to be this high. Thus it can be concluded that the cladding has pieked up carbon from the basemetal during welding of the cladding. These carbon levels place the cladding in the highly susceptible region for IGSCC [see Figure 21 FRACTURE MECIIANICS EVALUATION Even though we have found no evidence of penetration into the basemetal, in order to provide additional assurance of safe plant operation, Vermont Yankee had a fracture mechanics evaluation performed by SIA.
Fracture mechanies evaluations were performed to determine allowable daw depths as well as time to reach these allowable depths for Haws which might continue to pr.'pagate from possible through-clad cracks. The evaluations conservatively considered Haws in both the reador sessel head and the vessel wall adjacent to a steam dryer bracket.
=
Vessel llettd Allowable Haw sizes and the potential for flaw growth in the low alloy steel of the pressure vessel head at Vermont Yankee have been estimated using SIA's fracture mechanics software pc-CRACK. Recent data was used to derive a conservative crack growth rate for SCC-assisted cracking in the low alloy steel.
Stresses were obtained from the Chicago Bridge & tron (CB&I) stress report for the Vermont Yankee reactor vessel.
3
4 A 360' circumferential crack model in the cladding was assumed. Crack growth in the low alloy steel was assumed to begin immediately upon exposure to the water environment due to a through-clad crack.
Both of these assumptions are obviously conservative. A number of cases were run.
In the most limiting case, the applied K reaches a value of 63.25l 6.u b), where Ku = 200 ksidn from Figure A-4200-1 of ASME Section XI) at a depth of about 1.25 inebes. The crack growth analyses show that a crack could grow to a depth of 1.25 inch in approximately 56,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (about 6.5 years) for the base case. For the worst case, the time for a crack to reach that depth wouid be about 39,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />
( = 4.5 years). For locations where the residual stress is zero, the time for a Daw to reach one inch would be at least 77,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (= 8.8 years). Crack growth time to the allowable flaw size in the Gange area was on the order of 40 years.
As noted above, the assumptions used to construct these cases make the results extremely comervative.
In this analysis, cracks or other Daws that penetrated the stainless steel cladding are assumed to continue to propagate into the low alloy steel head immediately. This would not be e pected. Stress corrosion cracking can occur in low alloy steels in BWP environments, however, the crack growth rate expressions are derived from specimens loaded under conditions that are ideal for crack growth. Initiation times are typically not reported.
Of greate significance to this case is that a very small area of alloy steci will be exposed to the environment, unlike the crack growth tests where fairly substantial specimens are exposed. This area effect may be extremely significant relative to SCC susceptibility (particularly crack initiation) since cracking propens. v has been shown to be dependent upon sulfur content. In nigh purity environments, dissolution of mangwe sulfide inclusians produces an acidic sulfate enviranment at the metal surface which is conducive to crack growth. The dependence of SCC on both the number of inclusions and their shape demonstrates that cracking in high purity water is a function of the probability of the environment intersecting a manganese sulfide inclusion. Even a 360' tlaw through the entire thickness of the cladding would expose only a very small a ea oflow alloy steel. The probability of intersecting an inclusion is-highly unlikely.
= Reelon Near Steam Drver Bracket To determine the critical Oaw size in the reactor vessel wall, an analysis similar to that for the vessel head was conducted. A daw in excess of 1.5 inches was determined to be acceptable.
To determine IGSCC crack growth, the same conservative evaluat ion was conducted as noted above.
The crack growth analysis showed that the time for a crack to grow from the clad-to-base metal interface to 1.5 inches would exceed 40 years of operation. If the maximum computed clad stresses (for cold conditions) were used in the crack growth analysis, then the time to grow to 1.5 inches would approach 81,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> (approximately 9 yeats).
ASME CODE REOUIREMENTS The dryer support bracket was examimd per Section XI. The acceptance standards are provided in IWB-3520. The 1980 Edition of Section XI, with Winter 1980 Addenda, governs the Vermont Yankee ISI program.
That version of the Code states that the criteria for IWB-3520 are "in the course of preparation." The 1989 Edition of Section XI was consulted for guidance. The 1989 Code would reject " crack-like surface flaws on the welds joining the attachment to the vessel that exceed the allowable Haw standards of IWB-3510" (IWB-3520. l(a)). This visible indication exceeds the acceptance standards of Table IWB-3510.3. However, the location of the indication is adjacent to the blend radius of the attachment weld meets the vessel cladding. The blend radius does not add any strength to the attachment weld; its purpose is to act as a stress concentration mitigator. Thus, it could be judged that the Code does not apply to this location. A conservative approach is 4
l
- ' taken by assuming the Code, does apply. A supplemental examination (UT) was performed as allowed by IWB- l
- 3142.2. Since no relevant conditions were detected, the condition is deemed acceptable for continued service.
The inspection of the RPV head was being performed in response to the GE SIL 539, if the inspection was being performed as part of the inservice inspection program, the acceptance criteria would be obtained from IWB-3520. Since the 1980 Edition of the Code does not provide any standards for evaluation the 1989 Edition of the Code was again consulted for guidance. Paragraph IWB-3520.2 piovide acceptance standards.
lWB-3520.2 requires that relevant conditions be corrected in accordance with the requirements of IWB-3142. Paragraph IWB-3142.2 allows acceptance of relevant conditions by supplemental exnmination and directs the user to IWB-3200. Paragraph IWB 3200(b) states, " Visual examinations that reveal relevant conditions described in the standards of this Article may be supplemented by surface or volumetrie examinations to determine the extent of the unacceptable conditions and the need for corrective measures, repairs, analytical evaluation, or replacement."
For a given large population, a sample of acceptable measurements of the size taken by Vermont Yankee will accept that population with a high level of confidenee. The number of flaws evaluated provides a high level of confidence that there is no basemetal penetration and thus no corrective measures, repairs, analytical evaluations or replacements are required.
OIIIER CONSIDERATIONS ,
Consideration has been given to possible "spalling" of the cladding from the pressure vessel basemetal.
From an engineering perspective, there is negligible stress perpendicular to the basemetal/ clad interface to cause the IGSCC to " turn" and grow along the interface. It is now known that the dryer support bracket mdication is at least nine years old. That area was thoroughly examined nitrasonically to determine if any delamination was occurring. No evidence of any lack of bond was detected.
Consideration has also been given to the possibility of clad indications existing in other places of the reactor pressure vessel. The fracture mechanics evaluations were performed using end-of-life vessel fracture toughness properties for the vessel beltline region, thereby accounting for vessel embrittlement. As a point of interest, the end of life vessel fluence for Vermont Yankee is less than 2.3 x 10" n/ cia2 at the location of the peak
. iluence. This is a very low fluence compared to most reactor pressure vessels. Therefore, if similar flaws existed in the 14eltline region of the reactor pressure vessel the fracture mechanics evaluation bounds this condition.
Vermont Yankee also ultrasonically inspects the fou. feedwater inlet nozzles every refueling outage.
These nozzles are clad with stainless steel and are subject to thermal fatigue eyeling. No indication of any basemetal cracking has been detected.
CONCLUSIONS Based on the above observations and analysis, it is concluded that the following condition exists:
The RPV head and vessel clad indications and the indication at the dryer support bracket attachment weld result from 'GSCC susceptible clad ma:erial, sufficient stress .md an aggressive environment. The clad material is susceptible due to eladding diluthm during the weld metal application. This mechanism is the same for both the RPV head and vessel elad indications The sampling of indications that were evaluated using UT provide a high level of confidence that these cladding indications do not extend into the basemetal.
E
There is no evidence of cladding flaws extending into the basemetal. Growth of cladding flaws into the low alloy stepl basemetal is very unlikely. This is evidenced by the work performed by Common'vealth IIdison in their investigation of the Quad Cities event and supported by liPRI research on stress corrosion of low alloy steel.
Fracture mechanics evaluations demonstrate that in the worst ease three operating cycles would be required before any cladding flav /s would exceed the Code acceptable flaw size. These fracture mechanies evaluations have been performed using conservative assumptions. These evaluations show that full circumferential flaws in the reactor pressure vessel in excess of one inch deep would be acceptable under Code flaw evaluation rules. Conservative flaw growth calculations were performed.
EVE!RE INSPECTION Pld_NS Vermont Yankee is still evaluating the extent of future actions and will have these plans developed no later than 30 days prior to the 1993 refueling outage. As a minimum, Vermont Yankee will:
= re-inspect the dryer support bracket indication during the 1993 refueling outage and the following two examination periods.
= perform visual and UT inspectbn on the RPV head llange cladding during the 1993 refueling outage.
The inspection areas will t,e within the same areas as inspected in 1992.
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