Letter Sequence Response to RAI |
|---|
|
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MONTHYEARNG-08-0135, Request for NRC Approval of Alternative to Nozzle to Vessel Weld and Inner Radius Examinations2008-02-28028 February 2008
Request for NRC Approval of Alternative to Nozzle to Vessel Weld and Inner Radius Examinations
Project stage: Request
L-08-111, Inservice Inspection Program Relief Request IR-0542008-03-31031 March 2008
Inservice Inspection Program Relief Request IR-054
Project stage: Request
ML0819806282008-07-31031 July 2008
Request for Additional Information, Inservice Inspection Relief Request IR-054
Project stage: RAI
ML0820400462008-08-29029 August 2008
Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations Tac MD8193)
Project stage: Approval
L-08-270, Response to Request for Additional Information Regarding Relief Request IR-054, Revision 02008-09-17017 September 2008
Response to Request for Additional Information Regarding Relief Request IR-054, Revision 0
Project stage: Response to RAI
ML0831104542008-11-0606 November 2008
E-Mail Acceptance Review for Columbia Relief Request
Project stage: Acceptance Review
ML0831108332008-11-14014 November 2008
Request for Additional Information Related to Request for Relief 3ISI-09
Project stage: RAI
RS-08-156, Proposed Alternative to 10CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections2008-12-0303 December 2008
Proposed Alternative to 10CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections
Project stage: Request
ML0829607292008-12-29029 December 2008
Request for Relief Related to Inservice Inspection Relief Request IR-054
Project stage: Other
ML0902700232009-01-27027 January 2009
Acceptance Review of Proposed Alternative 10 CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections
Project stage: Acceptance Review
RS-09-044, Request for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations2009-03-13013 March 2009
Request for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations
Project stage: Request
ML0923003942009-08-24024 August 2009
Proposed Alternative to 10 CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections
Project stage: Other
ML0929404362009-11-0303 November 2009
Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations
Project stage: Other
ML1003500962010-02-0101 February 2010
Alternative Examination Requirements for Pilgrim Nozzle-to-Vessel Shell and Inner Radii Welds Using ASME Code Case N-702 an BWRVIP-108
Project stage: Other
ML1020202572010-07-13013 July 2010
Entergy Response to NRC Request for Additional Information Related to Pilgrim Relief Request (PRR)-20, Alternative Examination Requirements for Pilgrim Nozzle-to-Vessel Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108
Project stage: Response to RAI
ML1022901632010-08-25025 August 2010
Relief Request PRR-20, Alternative Examination Requirements for Nozzle-To-Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108 - Pilgrim Nuclear Power Station
Project stage: Other
JAFP-11-0112, Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP2011-10-0303 October 2011
Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP
Project stage: Request
2008-09-17
[Table View] |
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Text
I-g-- -- r FENOC f10 Perry Nuclear Power Station Center Road FirstEnergyNuclear OperatingCompany Perry,Ohio 44081 Mark B. Bezilla 440-280-5382 Vice President Fax: 440-280-8029 September 17,.2008 L-08-270 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant Docket No. 50-440, License, No. NPF-58 Response to Request for Additional Information Regarding Relief Request IR-054, Revision 0 (TAC No. MD8458)
By a letter dated July 31, 2008, the Nuclear Regulatory Commission (NRC) staff requested additional information related to Relief Request IR-054, Revision 0, which is a request for the Perry Nuclear Power Plant for relief from certain Inservice Inspection requirements associated with the implementation of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The response to the staffs Request for Additional Information (RAI) is attached.
On September 12, 2008, the NRC staff requested further information as part of the RAI response and it was agreed that a submittal date beyond 45 days for the response would be acceptable.
There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -
Fleet Licensing, at (330) 761-6071.
I declare under penalty of perjury that the foregoing is true and correct. Executed on September Iq , 2008.
Sincerely, 46c/7 4 (-K
Perry Nuclear Power Plant L-08-270 Page 2 of 2
Attachment:
Response to Request for Additional Information Related to Relief Request IR-054, Revision 0 cc: NRC Region IIIAdministrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board
Attachment L-08-270 Response to Request for Additional Information Related to Relief Request IR-054, Revision 0 Page 1 of 4 The following supplemental information is provided to respond to a Request for Additional Information (RAI) that was provided on July 31, 2008. The NRC question is repeated below, in bold, and is followed by the FirstEnergy Nuclear Operating Company (FENOC) response for the Perry Nuclear Power Plant (PNPP).
Background
The technical bases supporting your request for relief from the ASME Code, Section Xl examination requirements regarding RPV nozzle-to-vessel welds and nozzle inner radii and use of an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling- Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," for Perry are documented in BWR Vessel and Internals Project (BWRVIP) Report BWRVIP-108, "Technical Basis for the Reduction of Inspection Requirements for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii." The safety evaluation (SE) on BWRVIP-108 dated December 19, 2007, listed conditions for an applicant to demonstrate the plant-specific applicability of BWRVIP-108 to its plant. Your submittal indicated that all conditions specified in the SE are satisfied. One of the conditions is that the RPV heatup/cooldown rate is less than 115 F. However, the information from the NRC resident inspector at Perry indicated that, recently, Perry had more than two events with the heatup/cooldown rate exceeding 115° F.
Request For Additional Information Please discuss heatup/cooldown rate versus time information for these events and their frequency (how often does it occur). The staff plans to use this information to adjust the probabilistic fracture mechanics results reported in BWRVIP-108 to assess its impact.
Response to Request:
The table provided on the following pages lists those PNPP transient events that have occurred over the past ten years that have exceeded a heatup/cooldown rate of 115 F.
This table lists the date of each event, the temperature measurement location, the maximum heatup/cooldown temperature rate, and corresponding comments, which provide additional clarifying information about each event.
FENOC believes that the four transient events detailed in the following table do not invalidate the plant-specific applicability of the BWRVIP-108 report for the PNPP nor does this data conflict with the technical basis to incorporate Code Case N-702 provided in Relief Request IR-054, Revision 0.
Attachment L-08-270 Response to Request for Additional Information Related to Relief Request IR-054, Revision 0 Page 2 of 4 Key: CD = Cool Down HU = Heat Up BH = Bottom Head BHDN = Bottom Head Drain Maximum Date Event Measurement rate of Comment Location change in temperature 04/29/2001 CD Reactor -220 (0F/hr) Manual scram. Reactor Recirculation Recirculation pumps tripped.
Pipe Severe reactor recirculation pipe Loop A transients. Reactor pressure vessel temperature rates
+190 (°F/hr) measured from bulk saturation temperature remained within the 100°F/hr limit. Plant computer data used to determine maximum rates for reactor recirculation piping.
07/12/2001 CD BHDN -200 (0F/hr) Plant scram as a result of a loss of Feedwater. BHDN cool down. No other components were noted as affected. Reactor recirculation temperature did not oscillate. Reactor pressure vessel temperature rates measured from bulk saturation temperature remained within the 100°F/hr limit.
'Attachment L-08-270 Response to.Request for Additional Information Related to Relief Request IR-054, Revision 0 Page 3 of 4 Maximum Date Event Measurement rate of Comment Location change in temperature 12/15/2001 HU/CD BHDN -257 (0F/hr) Plant scram as a result of a loss of Feedwater. BH, BHDN, and reactor recirculation pipe BH -115 (°F/hr) experienced temperature changes in excess of 100 OF/hr.
Plant computer data used to Reactor +190(OF/hr) determine maximum rates for Recirculation reactor recirculation piping.
Pipe Reactor pressure vessel Loop B -177 (°F/hr) temperature rates measured from bulk saturation temperature remained within the 100°F/hr limit.
11/28/2007 CD/HU BHDN +259 (OF/hr) Plant scram as a result of a loss of Feedwater. Post scram, loss of Reactor Water CleanUp Reactor -268 (OF/hr) system and no vessel Recirculation recirculation flow - just natural pipe Loop A circulation. BHDN rate from surveillance data for one hour.
Recirculation temperatures Reactor +263 (°F/hr) -from plant computer data.
Recirculation Reactor pressure vessel pipe Loop B temperature rates measured from bulk saturation temperature remained within the 100°F/hr limit.
Additional discussion of this event is provided on the following page.
b -
Attachment L-08-270 Response to Request for Additional Information Related to Relief Request IR-054, Revision 0 Page 4 of 4 For the November 28, 2007 plant scram, the following was concluded after review and evaluation of plant operating data from this transient event:
- 1. The heatup/cooldown rate for the vessel head, bottom and shell flange were well within the 100 °F/hr limit of PNPP Technical Specification (TS) 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
- 2. The maximum heatup/cooldown rate for the bottom head drain and recirculation loop A/B nozzle exceeded 100 °F/hr. However, a report by a technical consultant, Stevenson &Associates, provided an evaluation demonstrating that the impact of heatup/cooldown rates in excess of those stated in PNPP TS 3.4.11 is acceptable and the usage factor for the affected reactor pressure vessel region and components is less than 1.00.
Therefore, for the November 28, 2007 plant scram, the reactor pressure vessel and all the affected components were determined to continue to meet the requirements of American Society of Mechanical Engineers (ASME) Code and are capable of performing their design function.
