ML090080799
| ML090080799 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/02/2008 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| Download: ML090080799 (9) | |
Text
PrairieIslandNPEm Resource From: Richard Plasse Sent: Tuesday, December 02, 2008 2:53 PM To: Eckholt, Gene F.
Subject:
PINGP RAI TLAA LETTER.doc Attachments: PINGP RAI TLAA LETTER.doc Latest draft, see new 4.3.1.4-2 writeup.
1
Hearing Identifier: Prairie_Island_NonPublic Email Number: 247 Mail Envelope Properties (Richard.Plasse@nrc.gov20081202145200)
Subject:
PINGP RAI TLAA LETTER.doc Sent Date: 12/2/2008 2:52:50 PM Received Date: 12/2/2008 2:52:00 PM From: Richard Plasse Created By: Richard.Plasse@nrc.gov Recipients:
"Eckholt, Gene F." <Gene.Eckholt@xenuclear.com>
Tracking Status: None Post Office:
Files Size Date & Time MESSAGE 42 12/2/2008 2:52:00 PM PINGP RAI TLAA LETTER.doc 90222 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
Mr. Michael D. Wadley Site Vice President Prairie Island Nuclear Generating Plant, Units 1 and 2 Northern States Power Company, Minnesota 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 & 2, LICENSE RENEWAL APPLICATION (TAC Nos. MD8513 AND MD8514)
Dear Mr. Wadley:
By letter dated April 11, 2008, Nuclear Management Company, LLC, now known as Northern States Power Company, Minnesota (NSPM), submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54) to renew the operating license for Prairie Island Nuclear Generating Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. Further requests for additional information may be issued in the future.
Items in the enclosure were discussed with Gene Eckholt, of your staff, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1427 or e-mail Richard.Plasse@nrc.gov.
Sincerely, Richard Plasse, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
As stated cc w/encl: See next page
Mr. Michael D. Wadley Site Vice President Prairie Island Nuclear Generating Plant, Units 1 and 2 Northern States Power Company, Minnesota 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 & 2, LICENSE RENEWAL APPLICATION (TAC Nos. MD8513 AND MD8514)
Dear Mr. Wadley:
By letter dated April 11, 2008, Nuclear Management Company, LLC, now known as Northern States Power Company, Minnesota (NSPM), submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54) to renew the operating license for Prairie Island Nuclear Generating Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. Further requests for additional information may be issued in the future.
Items in the enclosure were discussed with Gene Eckholt, of your staff, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1427 or e-mail Richard.Plasse@nrc.gov.
Sincerely, Richard Plasse, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION: See next page ADAMS Accession Number: ML083010561 OFFICE LA:DLR PM:RPB2:DLR BC:RPB2:DLR NAME I. King RPlasse DWrona DATE 11/25/08 11/21/08 11/ /08 OFFICIAL RECORD COPY
Letter to M. Wadley from R. Plasse, dated December XX, 2008 DISTRIBUTION:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 & 2, LICENSE RENEWAL APPLICATION (TAC NOS. MD8513 AND MD8514)
HARD COPY:
DLR RF E-MAIL:
PUBLIC B. Holian / S. Lee (RidsNrrDlr)
T. Combs, OCA T. Madden, OCA R. Shane, OCA B. Keeling, OCA D. Wrona (RidsNrrDlrRpb2)
R. Rikhoff (RidsNrrDlrRerb)
R. Plasse T. Wengert N. Goodman R. Skokowski K. Stoedter P. Zurawski OPA (RidsOpaMail)
I. Couret, OPA V. Mitlyng B. Mizuno, OGC OGC (RidsOGCMailRoom)
DLR/Rpb2
Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:
Peter M. Glass Assistant General Counsel Dennis L. Koehl Xcel Energy Services, Inc. Chief Nuclear Officer 414 Nicollet Mall (MP4) Northern States Power Company, Minneapolis, MN 55401 Minnesota 414 Nicollet Mall (MP4)
Manager, Regulatory Affairs Minneapolis, MN 55401 Prairie Island Nuclear Generating Plant Northern States Power Company, Joel P. Sorenson Minnesota Director, Site Operations 1717 Wakonade Drive East Prairie Island Nuclear Generating Plant Welch, MN 55089 Northern States Power Company, Minnesota Manager - Environmental Protection 1717 Wakonade Drive East Division Welch, MN 55089 Minnesota Attorney General=s Office 445 Minnesota St., Suite 900 St. Paul, MN 55101-2127 U.S. Nuclear Regulatory Commission Resident Inspector's Office 1719 Wakonade Drive East Welch, MN 55089-9642 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 85 7th Place East, Suite 500 St. Paul, MN 55101-2198 Tribal Council Prairie Island Indian Community ATTN: Environmental Department 5636 Sturgeon Lake Road Welch, MN 55089 Charles R. Bomberger Vice President Nuclear Projects 414 Nicollet Mall, (MP4)
Minneapolis, MN 55401
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION RAI 4.7.4-1 License renewal application (LRA) Section 4.7.4 indicates that the polar cranes, auxiliary building cranes, the turbine building cranes, and spent fuel cranes will be used less than 20,000 lifting cycles over 60 years.
Provide your estimate for the number of lifting cycles that have occurred in each of the stated cranes and your 60-year lifting cycle projections for these cranes.
RAI 4.3.1.3-1 To project the cycles and cumulative usage factor (CUF) to the extended period of operation, it is necessary to have the base line information. Before the dates of issuance of NRC Bulletins 88-08 and 88-11, there wasnt any data collected regarding stratification and insurge/outsurge events for pressurizer surge lines.
Please discuss how you reconstructed the cycles that occurred prior to December 20, 1988 (the date of issuance for Bulletin 88-11), for the pressurizer surge line stratification events and insurge/outsurge events. Please provide the date when the events tracking began and discuss how the tracked data was used to achieve the 60-year cycle projection.
RAI 4.3.3-1 It is known that some of the reactor coolant system piping charging nozzles are designed and manufactured with a thermal sleeve and some are not.
Please confirm whether or not a thermal sleeve exists in the charging nozzle, and confirm that the stress and fatigue analyses were performed based on the actual geometric conditions of the charging nozzle in use.
RAI 4.3.1.1-1 Some of the reported CUF results were generated by FatiguePro software. However, FatiguePro is not endorsed by NRC staff, since FatiguePro does not produce all six individual components of a transient stress tensor (Sxx, Syy, Szz, Sxy, Syz, Szx) to support the American Society of Mechanical Engineers (ASME)Section III fatigue analysis method. FatiguePro produces only one stress component and uses that single stress component for fatigue evaluation.
Please identify from Tables 4.3-2 through 4.3-8, if any, those items whose CUF values were calculated without using every individual component of the transient stress tensors. The items that are identified must be reevaluated in accordance with the ASME Section III guidelines. In addition, the staff noticed multiple occurrences of the terminology stress-based monitoring in the body of the LRA.
If the plant does not have appropriate stress monitoring capability, use of such a terminology would be misleading. Please make appropriate corrections.
ENCLOSURE
RAI 4.3.1.1-2 Tables 4.3-2 through 4.3-8 show the CUF results. However, it is not clear whether the CUF values listed in these tables represent design basis CUF values or 60-year projected CUF values. Please clarify whether the CUF values provided in LRA Table 4.3-2 through 4.3-8 represent design basis (40 year) CUF values or 60-year projected CUF values.
RAI 4.3.1.2-1 The design cycles and the projected 60-year cycles for Prairie Island Nuclear Generating Plants (PINGPs) design basis transients are shown in Table 4.3-1 of the LRA. The staff has noted that the paragraph under LRA Table 4.3-4 of the LRA, which discusses the plant loading and unloading transient cycles, is vague. In the targeted paragraph, an unrelated/unrecognized number, 1835, which is not shown in Table 4.3-1, is suddenly brought up. Clarify and indicate the source of this 1835 number and clarify whether this value represents the new design basis cycle values for the baffle bolts.
RAI 4.3.1.4-1 Table 4.3-6 of the LRA shows the CUF values at certain components in the steam generators. It is understood that the CUF values for Unit 1, reflect steam generator replacement (occurred in November 2004), therefore the CUF at the corresponding components/locations of Units 1 and 2, are expected to be somewhat different. However, for the primary inlet nozzle and primary outlet nozzle, the differences between Unit 1 and Unit 2, are significant. Specifically, at these two nozzles, the CUF is as high as 0.880 for the original steam generator (Unit 2) but it is dropped to near zero, 0.007, for the replaced one (Unit 1). Explain how this is possible. Specify the analytical tool used along with a discussion concerning differences in the input data.
RAI 4.3.1.4-2 (a) In Section 4.3.1.4 of the LRA, the applicant discusses fracture mechanics analyses for the steam generator feedwater nozzle and performed fracture mechanics analyses to justify leaving the flaws in the thermal sleeves in the as-found condition. Describe both the loading conditions and geometrical conditions used in the fracture analyses:
- For the loading conditions, indicate whether the transients used in the fracture mechanics analyses are the same as those shown in Table 4.3-1 of the LRA and if different explain why. Also, indicate whether the nozzle loads (axial force, bending and torsion) have been taken into account in these fracture mechanics analyses.
- For the geometrical conditions, provide a sketch showing the values for the main dimensions such as inside radius, outside radius, and thickness, as well as the flaw location, flaw shape, flaw aspect ratio, and flaw orientation of the postulated crack.
Provide the values of these parameters/dimensions used to determine the reported critical flaw depth (1.8 inches), including the value of fracture toughness, KIC. For the fatigue crack growth analyses performed, indicate the fatigue crack growth law used.
(c) In subsection labeled SG Tube Fatigue for Units 1 and 2, under Section 4.3.1.4 of the LRA, compliance with 10 CFR 54.21(c)(1)(ii) is exclusively for Unit 2. Provide your basis for omitting a similar statement for the Unit 1 steam generators.
RAI 4.3.1.5-1 In Section 4.3.1.5 of the LRA, the applicant states that Exemption from fatigue evaluation was justified for the casing feet, casing nozzle, and upper and lower seal housings and bolts, without supporting this statement with a regulatory basis. Provide your technical and regulatory basis for exempting the reactor coolant pump casing feet, casing nozzle, and upper and lower seal housings and bolts from being within the scope of an ASME Section III CUF analysis.
RAI 4.3.3-1 LRA Section 4.3.3 discusses environmentally assisted fatigue (GSI-190) and provides values of the environmentally assisted fatigue correction factors, Fen, applicable to PINGP. In the calculation of Fen for the low alloy steels, the applicant assumes that the dissolved oxygen level for pressurized-water reactor plants is below 0.05 ppm at temperatures above 150 oC. This assumption in effect takes away the dependency of Fen on strain rate and sulfur content and also significantly weakens the dependency of Fen on temperature. Provide your basis for using this assumption.
RAI AMP B2.1.3-2 During review of LRA Section B2.1.3, Operating Experience the staff noted that there were two instances of cracking due to intergranular stress corrosion cracking (IGSCC) of Safety Injection Accumulator Tanks. Although these cracks were repaired and the In-service Inspection Program was augmented to include additional inspections, such as ultrasonic testing and liquid penetrant inspection, the applicant did not indicate if new activities were implemented to improve and monitor the environment to preclude IGSCC. Please provide details of any preventive activities to preclude IGSCC.