ML18039A422
| ML18039A422 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/01/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18039A421 | List: |
| References | |
| NUDOCS 9807090300 | |
| Download: ML18039A422 (12) | |
Text
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VAL A 0 8
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D ELIEF RE E
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PLANT 1.0 IN D
N According to the provision of'10 CFR 50.55a(a)(1), "structures, systems, and components...[at nuclear power generating facilities are required to be]...designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed." According to 10 CFR.50.55a(a)(2), "systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code spe'ciflied in...this section."
However, the Director of the Office of,Nuclear Reactor Regulation may, grant relief of the ASME Code requirements if, pursuant to 10 CFR 50.55a(a)(3)(i), "proposed alternatives would provide an acceptable level of.quality and:safety."
By letter dated February 17, 1998as supplemented by letter;.dated June 12, 1998, the licensee submitted revised relief request 3-ISI-1'. TVAhad previously submitted the results of its augmented examination of the Browns Ferry Unit 3 (BF3) reactor pressure vessel welds which was conducted in accordance with 10 CFR 50.55a(g)(6)(ii)(A)(2)',. This regulation. requires that licensees perform volumetric examination of "essentially,100 percent" of the reactor pressure vessel (RPV) pressure-retaining shell welds to the extent:practical within the limitations of design and geometry.
The examination results revealed fifteen flaws in circumferentially oriented welds that exceeded the acceptance criteria in American'Society, of Mechanical Engineers. (ASME)Section XI, Paragraph IWB-3500. These flaws were evaluated in accordance with ASME Section XI, Paragraph IWB-3600, and found to be. acceptable for. continued service.
The staffs detailed safety evaluation report (SER) for the weld flaw evaluation was issued by letter dated November 8, 1995.
ASME Code,Section XI, Paragraph IWB-2420(b) requires that flaws identified in components that are found acceptable for continued service shall be reexamined during the next three inspection periods.
The original relief request 3-ISI-1 sought relief from the three successive ATTACHMENT1 Results of the augmented examination of Browns Ferry Unit 3 RPV welds was submitted by letter dated March 6, 1995.
9807090300 980701 PDR ADOCK 05000296 P
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2 reexaminations of the Unit 3 RPV circumferential weld flaws. By'letter dated July 8, 1997, the NRC staff denied TVA's original request.
The revised request seeks relief.for one operating cycle from performing the first reexamination:
The.'licensee proposes to perform the first reexamination during the cycle 9 outage (scheduled,to begin March 2000) as opposed'to the cycle 8 outage (scheduled to begin October 1998).
plicable components are, the BF3 RPV welds listed bel Vessel to Flange Weld - C-5-FLG Vessel Shell Circumferential'Welds - C-2-3 C-3-4; C-4-5 2.0 BAS F
THE REL E E
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The ap ow'989 Edition, no Addenda, of, the ASME Code:
Section XI, Paragraph IWB-2420(b) requires that licensee's evaluate flaws or relevant conditions in accordance with IWB-3132.4 or IWB-3142.4, respectively, and ifthe component qualifies as acceptable for continued service, the areas containing such flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedules of the inspection programs of IWB-2410.
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'e ue Relief is requested from performing the successive volumetric examinations of the RPV areas were flaws were found during the BF3 Unit 3 extended outage.
These flaws were found in circumferential welds. The applicable time period for which relief is requested is one operating cycle.
Relief is requested on the basis that, 1) the flaws are subsurface and result from fabrication of the vessel, 2) a General Electric (GE) flaw evaluation shows that the maximum indication depths (2a) willnot exceed the ASME Code-allowable flaw depths during the intended service life of the vessel., 3) the GERIS 2000 inspection equipment is unavailable due to other contractual obligations, and 4) industry initiatives and information support granting the relief.
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er The'licensee proposes to perform the first reexamination required by IWB-2420(b) during the cycle 9 outage (scheduled to begin. March 2000) as opposed to the cycle 8 outage (scheduled to begin October 1998).
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The licensee contracted with GE to perform the RPV shell welds augmented examination at BF3 during the cycle 5 extended outage (Fall 1993). The GE.Remote Inspection System (GERIS) 2000 was the ultrasonic (UT) equipment used to conduct the inspection.
In-addition, GE manually examined selected areas from the outside of the RPV in order to maximize the percentage of weld volume examined. A total of five circumferential and fifteen axial welds were examined.
The examination results revealed fifteen flaws in circumferentially oriented welds that exceeded the acceptance criteria'in ASME Section XI, Paragraph IWB-3500. Four of the RPV shell welds had a total of ten indications. that, exceeded the allowable standards. of the A'SME Code Section XI, IWB-3500. The indications were:located in weld'C-2-'3;,'C-3-4'; C-4-5, and V-4-B.
One indication was located in both weld,'V-4-'B and.,C-3-4,at the intersecting weld joint. The remaining five indications that exceeded. the IWB-',3500,standards were. in weld C-5-FLG. This weld was not part of the augmented examination,, but was evaluated along with the ten indications in'the.above mentioned'welds.'ll:of the indications were'located in the vessel flange welds and non-beltline region welds.;
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The measured depth (2a).of indication number 12-116 in weld C-'34 is 0.62:inches.
This is the largest UT measured depth. of,;the fifteen flaws.'The'measured length for indication 12-116 is 0.75 inches.
Indication number'12-148, also in weldiC-3-4; has the longest measured length of 2.75 inches.
Its measured "depth,is0:51.1 inches;.'These two'indications were found in the non-beltline circumferential welds. 'The licensee'characterized all,of the indications as subsurface flaws and as volumetric, anomalies caused:by,fabrication.
They were not previously detected with UT at the time,of fabrication."
The licensee. performed.a flaw evaluation in.accordance with IWB-3600 (1986 Edition) acceptance criteria. The flaw evaluation was'based on comparing the indications to the allowable flaw sizes that were developed in a bounding analysis performed by GE. The analysis developed the allowable flaw.size for an irradiation level and fatigue crack growth corresponding to 12 effective full power years (EFPY). The appropriate loadings were considered, and the upper bound of the allowable flaw size was established by ASME Section III!requirements for primary local stress which states that the maximum primary membrane, stress cannot exceed 1.5S
. GE limited the flaws to be within 1/3 of the thickness of the base metal. The'lower bound of the allowable flaw size was established by. the
.acceptance criteria of IWB-3500.
The staffs SER dated November 8, 1995 concluded that 1) the indications are within the ASME
- Code, IWB-3600 acceptance criteria, 2) BF3 is acceptable for continued operation for at'least 12 EFPY,, 3) the licensee is required to submit an analysis'to justify continued operation beyond 12 EFPY, and 4) TVAis required to reexamine the indications ig the next three inspection periods in accordance with IWB-2420(b).
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C; GE performed a flaw evaluation. for.the, BF3:RPV in accordance,with'Section XI of the ASME Code for.all axial and circumferential welds'in.the vessel shell, top head,'ottom head, and flange regions.
The results. were documented as:a series of "flawallowable curves" in the GE report, GENE-523-120-'0992.
This report is,'Reference
- 1. to Attachment'3 of the current submittal, and forms.the basis:for. the revised.'analysis. which,extends the'..flaw handbook results from 12 effective full power years (EFPY) to the inten'ded service life of the vessel.
The staff examined all three aspects of the'flaw evalu'ation: the calculation, for the:applied stress intensity factor, K,,(driving force), the calculatior'i for the fracture. toughness, K, (resistance),
and the acceptance criteria (relationship between, driving force and'resistance).
The staff agrees that using hydrotest and boltup conditions as the limiting load condition;is appropriate, because this load was identified as the most limiting one through, previous analyses by GE, and was approved by, the staff and used in all previous flaw:evaluations.by licensees for other BWR vessels.
Specifically, this report considered (1).clad residual stress,.(2) bolt preload stress, (3) pressure stress, and (4) weld residual stress in the applied K calculation In estimating the resistance to fracture, the staff verified.that the report used the Kcurve of Section XI of the ASME Code based on the adjusted'reference temperature (RT>>,) of each weld. The submittal dated November 24, 1997, indicated that the flaw allowable curves are based on irradiation and the associated leak test temperatures at 12 EFPY. This statement prompted the staff to question the applicability of these curves at other EFPY. The licensee
. provided its response dated June 12,.1998, to the staffs request for additional information (RAI), and demonstrated'that operation of the BF3 RPV at the limits validated by the 32 EFPY pressure-temperature curves would compensate for any expected shift in the RT>>~ for all of the vessel welds. The acceptance criterion used by the licensee is K,~>> = K,J(10)'". This is in accordance with IWB-3612 of Section XI'ofthe ASME Code.
Another staff concern is the adequacy of using the applied K formula for a semi-infinite crack in an infinite sheet to the present, case of a finite crack in a finite sheet.
The qualitative explanation in,the licensee's.,response to the staff's RAI appears to be reasonable.
- However, no quantitative assessment was provided.
The staff confirmed that the analysis performed for the evaluation is consistent with the ASME Code methodology, and that the indications are within the allowable flaw size. The staff accepts the submittal-because of the large margin between the flaw sizes of the 1993 UT indications and the proposed'limit for the flaw allowable curves for each weld considered in the analysis.
This safety evaluation should not be interpreted as acceptance of the report, GENE-523-120-0992, by the staff..Without a quantitative assessment on using the closed-form applied K solution for a.semi-infinite crack in an infinite sheet to the RPV, the staff will'review future submittals using GENE-523-120-0992 on a case by case basis.
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4.0 The staff has reviewed the licensee's submittal and concludes that the analysis is acceptable for the following reasons:
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The flaws are subsurface.
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The GE flaw evaluation shows that the maximum indication depths (2a) will not exceed the ASME Code-allowable flaw depths during the intended service life of the vessel.
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The appropriate tooling and equipment are unavailable for use during the cycle 8 outage.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), relief is.granted for one operating cycle.
5.0 1)
ASME Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, American Society of;Mechanical Engineers, 1989 Edition, Paragraph IWB 3640.
2) 3)
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ASME Boiler and Pressure'Vessel Code.Section III, Rules for the Construction of, Nuclear Power Plant Components, American'Society of Mechanical Engineers, 1989 Edition.
Letter to O. D. Kingsley (TVA)from J. F; Williams (USNRC)
Subject:
."Browns Ferry Nuclear Plant Unit 3 - Reactor Vessel Weld Flaw,'Evaluation (TAC M93759)," November 8, 1995.
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Letter to O. D. Kingsley (TVA)from F. J. Hebdon (USNRC)
Subject:
"Relief Request - Browns Ferry Nuclear,Plant Unit'3 Relief, Request 3-ISI-1 (TAC M97805),"
July 8, 1997.
5)
Letter to USNRC Document Control Desk from T. E. Abney (TVA)
Subject:
"Browns Ferry Nuclear Plant (BFN) - Unit 3 Revised Relief Request 3-ISI-1 Regarding Reactor Pressure Vessel (RPV) Shell Welds Augmented and American Society of Mechanical Engineers (ASME)Section XI Inspections and TVA's Reply to NRC's Letter ti TVA Dated July 8, 1997." February 17, 1998.
6)
Letter to O. J. Zeringue (TVA)from A. W. De Agazio (USNRC)'
Subject:
"Request for Additional Information - Browns Ferry Nuclear Plant, Unit 3 Revised Relief Request 3-ISI-1 (TAC MA1153)," May 19, 1998.
7)
GE Report GENE-523-B1301869-129, "Extension of Unit 3 Vessel Flaw Handbook Results to 40 years," November 21, 1997
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AL S
ARY REV EWA TIVITIES The staff reviewed the licensee's revised relief request 3-ISI-1 for the Browns Ferry Unit 3 reactor pressure vessel.
The staff considered the bases for the requested relief, and confirmed that the analysis performed for the flaw evaluation is consistent with the ASME Code methodology, and that the indications. are within the allowable flaw size.
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IA'i The licensee used good quality control in preparing the submittal which was relatively thorough.
The additional information, needed to complete the review'was submitted in a timely manner, and adequately addressed the outstanding concerns,.with the exception of a quantitative assessment on using the closed-'from appliediK solution for a semi-infinite crack in an infinite sheet to the RPV flaw evaluation.
Author:
A.D. Lee (301) 415-2735 ATTACHMENT2
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