ML20151L615
| ML20151L615 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/28/1997 |
| From: | Jain S DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M98137, TAC-M98138, NUDOCS 9708070084 | |
| Download: ML20151L615 (4) | |
Text
-
t Beaver Valley Power Station Shippingport. FM 15077 0004 July 28, 1997 c
Nuclear Power Division U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Response to Request for Additional Information Technical Specification Change for CE SG Tube Sleeves (TAC Nos. M98137 and M98138) i Attached is our response to an NRC request for additional information dated June 23,1997, concerning a technical specification change that would allow the use of steam generator tube sleeves designed by ABB/ Combustion Engineering that was submitted by our letter dated March 10,1997. This response confinns the existing tube plugging / repair limit, agrees with NRR issuance of a license condition to perform post 4
weld heat treatment and addresses sleeve inspection criteria. The attachment lists each NRC item followed by our response.
The response to item I states that a validation of the Unit 2 steam generator tube repair limit to the methodology detailed in Regulatory Guide 1.121 will be performed.
We believe this issue bears no relation with the NRC review of this proposed technical specification change for Unit 1.
We request that the NRC approve the subject amendment for implementation during the scheduled September 1997 refueling outage.
i I
I Sincerely, gd[
Sushil C. Jain
)
l c:
Mr. D. M. Kern, Sr. Resident inspector Mr. H. J. Miller, NRC Region I Administrator DMM Mr. D. S. Brinkman, Sr. Project Manager 0UA.LITV ll!!!!!f!!!!f!I! h!!,!!!!!f!!!!f!! !!
9700070084 970728 PDR ADOCK 05000334 P
ATTACHMENT Beaver Valley Power Station, Unit No. I and No. 2 Response to Request for Additional Information Technical Specification Change for CE SG Tube Sleeves Item 1 l
The licensee should confinn that the existing TS plugging / repair limit for tubes (40%) is correct or provide an appropriate revision. Historically, some TSs used certified material test report (CMTR) values for tube material strength instead of American Society of Mechanical Engineers (ASME) Code minimums. The plugging / repair limit should use the ASME Code minimum for material strength in the calculation of Code allowable flaw depth minus NDE uncertainty (20%). The staff notes that in the amendment submittal the plugging limit for the CE sleeves (32%) is calculated in this desired manner.
Response 1
~
As stated in the NRC Safety Evaluation Report for Beaver Valley Unit 1, Supplement 3, dated March 19,1976, Section 5.2, REACTOR COOLANT SYSTEM, the disposition of the steam generator tube integrity issue for BVPS Unit I was encompassed by the proceedings and testimony provided for Northern States Power Company's Prairie Island Generating Plant, Units 1 and 2, Dockets 50-282 and 50-306 before the Atomic Safety and Licensing Appeal Board. The testimony of James Knight in January 1975 before the Atomic Safety and Licensing Board summarized the bases for the 40% steam generator tube repair limit. Mr. Knight's testimony referenced Westinghouse Topical Report WCAP-7832 entitled " Evaluation of Steam Generator Tube, Tube Sheet and Divider Plate Under Combined LOCA Plus SSE Conditions" which provided the detailed bases for the proposed plugging criterion. Knight's testimony concludes that the minimum acceptable tube wall thickness that must be maintained during operation of the reactors is 0.025 inch. This conclusion is based upon the analyses documented in WCAP-7832. He further states in his testimony that a corrosion allowance must be added to the 0.025 inch minimum wall thickness to establish the unacceptable defects which require tube plugging. This corrosion allowance was subsequently discussed in the testimony of Raymond Maccary before the Atomic Safety and Licensing Appeal Board in January 1976, again addressing the steam generator tube integrity issue for Prairie Island Unit I and 2 steam generators. Per Mr. Maccary's testimony, a corrosion allowance of 10% was established. Thus, the steam generator tube plugging limit was established as 40% of the tube wall thickness. Per Beaver Valley Unit l's SER, the testimony of both Knight and Maccary, as well as the referenced WCAP-7832 document the bases for the Beaver Valley Unit 140% steam generator tube repair limit. The material property values utilized in WCAP-7832 were ASME Code minimums.
[,<*,,
l^
Attachment.
Re~sponse to Request for Additional Information Technical Specification Change for CE SG Tube Sleeves Page 2 i
Although a corrosion allowance was specified, a NDE uncertainty was not included in the steam generator tube repair limit. Since the design bases for the parent steam generator tube repair limit does not include NDE uncertainty and since the supporting I
documentation adequately describes the inherent margins of safety associated with the
'40% steam generator tube repair limit, inclusion of a 20% NDE uncertainty in the parent l
tube repair limit is inappropriate. Additionally, in WCAP-7832-A dated April 1978, the l
minimum uniform wall thickness for Series 51 steam generator tubes (0.875 inch l
nominal diameter and 0.050 inch nominal wall thickness) is 0.021 inches which would provide additional margin for steam generator tube integrity above that discussed in l
Knight's and Maccary's testimony. Therefore, it is concluded that the existing Unit 1 technical specification repair limit of 40% for steam generator parent tubes is acceptable.
- For Beaver Valley Unit 2, UFSAR Table 5.4-4 states that the 40% tube plugging limit i
will be determined in accordance with Regulatory Guide 1.121. DLC is in the process of validating the Beaver Valley Unit 2 steam generator tube repair limit to the methodology 4
detailed in RG 1.121 and will provide these results to the NRC by November 28,1997.
Item 2 i
)
The licensee committed to perform post weld heat treatment (PWHT) of the freespan l
This commitment is congruent with the staff position on PWHT.
i However, there is. no accompanying TS wording reflecting this commitment.
Consequently, the licensee should provide suitable TS wording to include PWHT in the sleeving process or the Office of Nuclear Reactor Regulation (NRR) should issue the amendment with a license condition to perform PWHT. Should the licensee decide to omit PWHT, then supporting technical justification should be submitted.
Such l
justification should address issues including: projected impact on service life of the sleeve joints; anticipated plans for future repairs to the sleeve joints and/or steam generator replacement; and whether the omission of PWHT would result in a new or accelerated mode of degradation.
Response 2 DLC reiterates its commitment to perform PWHT when installing the ABB-CE TIG welded sleeves. Similar to the previous Beaver Valley TS amendment regarding laser welded sleeves, rather than revising the TS wording, it is more desirab!a for NRR to issue the TS amendment with a license condition to perform PWHT.
y
..*i,
)
. Attachment Response to Request for Ad_ditional Information Technical Specification Change for CE SG Tube Sleeves i:
Page 3 Item 3 The staff believes that the licensee should adopt inspection and expansion criteria l
consistent with that of other facilities licensed to install sleeves. As a minimum, the initial sample selected at each inspection should include 20 percent of all sleeved tubes.
Please provide appropriate wording and a table in the TS reflecting this criteria. For reference, a previously adopted table incorporated in the TSs of another licensee is attached.
Response 3 i
DLC has proactively implemented the inspection sampling requirements of the EPRI l
PWR Steam Generator Examination Guidelines which would result in a 20% initial sample of tubes repaired by sleeving if sleeves had been installed. While DLC intends to i
continue to utilize the aforementioned EPRI Examination Guidelines, we do not believe it is either necessary or appropriate to incorporate increased inspection sampling requirements in the technical specifications for repaired tubes beyond those presently established in the technical specifications for the parent tubes or those previously 6
approved for other sleeve designs.
h I