ML18039A486
| ML18039A486 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/17/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18039A485 | List: |
| References | |
| NUDOCS 9808240120 | |
| Download: ML18039A486 (12) | |
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UNITED STATES NUCL'EAR REGULATORY COMIVIISSION WASHINGTON, D.C. 20555-0001 N
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0 50- 96 Title 10 of the Code of Federal Regulations (CFR); Part 50.55a (10 CFR 50.55a), requires, in part, that in service inspection (ISI) of certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class I, 2, and 3 components be performed in accordance with Section XI of the ASME Code applicable Edition and Addenda, except where specific written relief has been granted by the Commission pursuant to 10 CFR,50.55a(g)(6)(i).
Title 10 CFR, Part 50.55a(a)(1) requires that structures, systems, and components at nuclear power generating facilities shall be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance. of the safety function to be performed.
Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Code as specified. in 10 CFR 50.55a(a)(2).
Proposed:alternatives to the ASME Code requirements may be authorized by the Director of the Office of Nuclear Reactor Regulation pursuant to 10 CFR 50.55a(a)(3) when the licensee demonstrates that the proposed alternatives provide an acceptable level of quality and safety, or that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
By letter dated February 17, 1998, as supplemented by letters dated June 12 and July 31, 1998, the Tennessee Valley Authority (TVA).submitted Revised Relief Request 3-ISI-1. TVA.had previously submitted the results of its augmented examination of the Browns Ferry Unit 3 (BFN-3) 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.
TVAsubmitted the results of the augmented examination of Browns Ferry Unit 3 reactor pressure vessel by letter dated March 6, 1995.
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The examination results revealed fifteen flaws in circumferentially-oriented welds that exceeded the acceptance criteria in ASME Code,Section XI, Paragraph IWB-3500. These flaws were evaluated in accordance with ASME Code,Section XI, Paragraph IWB-3600, and found to be acceptable for continued service.
The U.S. Nuclear Regulatory Commission (NRC) staff's detailed safety evaluation (SE) 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 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.
TVAproposes 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).
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The applicable components are the BFN-3 RPV welds listed below:
- Vessel.to Flange Weld - C-5-FLG
- Vessel Shell Circumferential Welds - C-2-3; C-3-4', C-4-5 1989 Edition, no Addenda, of the ASME Code:
Section XI, Paragraph IWB-2420(b) require's that licensee's evaluate flaws or relevant conditions in accordance with IWB-3132.4 or IWB-3142.4, respectively, and ifthe component qualiTies 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.
2.3 Relief is requested from performing the. successive volumetric examinations of the RPV areas
.where flaws.were found during the BFN-3 Unit 3 extended outage.
These flaws were found in circumferential welds. The applicable time period forwhich 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,
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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.
,2;5 TVAproposes to perform the first reexamination required by IWB-2420(b) during the Cycle 9 refueling outage scheduled to begin March 2000 as opposed to the Cycle 8 outage scheduled to begin October 1998.
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99 'S TVAcontracted with the GE to perform the RPV shell welds augmented examination at BFN-3 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 Code Section XI, Paragraph IWB-3500. Four of the RPV shell welds had a total of ten indications that exceeded the allowable standards of the ASME Code Section.XI, IWB-3500. The indications were located in welds C-2-3, C-3-4, C-4-5, and VA-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. Allof the indications were located in the vessel flange welds and non-beltline region welds.
The measured depth, 2a, of indication number 12-116 in weld C-3-4 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 weld C-3-4, has the longest measured length of 2.75 inches.
Its measured depth is 0.511 inches. These.two indications were found in the non-beltline circumferential welds. TVAcharacterized 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.
TVAperformed 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 Code Section III requirements for primary local
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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 NRC staff's SE dated November 8, 1995, concluded that (1) the indications are within the ASME Code, IWB-3600 acceptance criteria, (2) the BFN-3 RPV is acceptable for continued operation for at least 12 EFPY, (3) TVAis required to submit an analysis to justify continued operation beyond 12 EFPY, and (4) TVAis required to reexamine the indications in the next three inspection periods in accordance with IWB-2420(b).
GE performed a flaw evaluation for the BFN-3 RPV in accordance with Section XI of the ASME Code for all axial and circumferential welds in the vessel shell, top head, bottom 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 EFPY to the intended service life of the vessel.
The NRC staff examined all three aspects of the flaw evaluation: the calculation for the applied stress intensity factor, K (driving force), the calculation for the fracture toughness, Ki, (resistance),
and the acceptance criteria (relationship between driving force and resistance).
The NRC staff, agrees that using hydro test 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 boiling water reactor 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. TVAprovided its June 12, 1998, response to the staffs request for additional information (RAI), and demonstrated that operation of the BFN-3 RPV at the limits validated by the 32 EFPY pressure-temperature curves would compensate for any expected shift in the RTQ>> for all of the vessel welds. The acceptance criterion used by TVAis~~ = Kg(10)~. This is in accordance with IWB-3612 of Section XI of the 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 TVA's response to the staff's RAI appears to be reasonable.
However, no quantitative assessment was provided.
The NRC 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 limitfor 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 willreview future submittals using GENE-523-120-0992 on a case-by-case basis.
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The staff has reviewed TVA's submittal and concludes that the analysis is acceptable for the following reasons:
~ The flaws are subsurface.
~ 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.
~ The appropriate tooling and equipment are unavailable for use during the Cycle 8 outage.
Therefore, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i), the alternative is authorized for one operating cycle in that it provides an acceptable level of quality and safety.
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- 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) ASME Boiler and Pressure Vessel Code..Section III, Rules for the Construction of Nuclear Power Plant Components, American Society of Mechanical Engineers, 1989 Edition.
3)
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.
4)
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 to TVA Dated July 8, 1997."
February 17, 1998.
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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 Principal contributor.
A. Lee Date:
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