LR-N04-0002, Request for Additional Information Response - Hope Creek Generating Station Relief Request HC-RR-B12

Request for Additional Information Response - Hope Creek Generating Station Relief Request HC-RR-B12

ML040420119
Person / Time
Site:	Hope Creek
Issue date:	01/30/2004
From:	Mannon S Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N04-0002, TAC MB8407
Download: ML040420119 (5)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 JAN 3 0 2004 OPSEG LR-N04-0002 NuclearLLC United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 INSERVICE INSPECTION PROGRAM HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NOS.60-354

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RESPONSE -

HOPE CREEK GENERATING STATION RELIEF REQUEST HC-RR-B12 (TAC NO. MB8407)

By letter dated April 14, 2003, PSEG Nuclear LLC (PSEG) submitted a request for relief from the required reactor pressure vessel volumetric examination required by the American Society of Mechanical Engineers Code,Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.100. The relief was requested pursuant to Title 10 of Code of Federal Regulations Section 50.55e(a)(3)(ii).

The Nuclear Regulatory Commission (NRC) requested additional information be provided in response to their June 26, 2003 letter. By letter dated September 3, 2003 PSEG submitted the requested additional information.

On November 19, 2003 PSEG was contacted by the Hope Creek Project Manager regarding some additional requests for information. These were discussed with the NRC staff on December 18, 2003. The following is being submitted in response to the additional request for information.

Should you have any additional questions, please contact Mr. Michael Mosier at 856-339-5434.

Sincerely, Steven Mann Manager- Nuclear Safety and Licensing 95-2168 REV. 7/99

Document Control Desk LR-N04-0002 JAN 3 ° 2004 C Mr. H. Miller Regional Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. Nuclear Regulatory Commission ATTN: Mr. John Boska Licensing Project Manager - Hope Creek Mail Stop 08B1 Washington DC 20555-001 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625 2

Document Control Desk LR-N04-0002 REQUEST FOR ADDITIONAL INFORMATION HOPE CREEK GENERATING STATION By letter dated April 14, 2003, PSEG Nuclear LLC (PSEG) submitted a request for relief from the required volumetric examination required by the American Society of Mechanical Engineers (ASME) Code, Section Xl, Table IWB-2500-1, Examination Category B-D, Item B3.100. The relief was requested pursuant to Title 10 of Code of Federal Regulations Section 50.55e(a)(3)(ii).

The Nuclear Regulatory Commission (NRC) requested additional information be provided in response to their June 26, 2003 letter. By letter dated September 3, 2003 PSEG submitted the requested additional information.

On November 19, 2003 PSEG was contacted by the Hope Creek Project Manager regarding some additional requests for information. These were discussed with the NRC staff on December 18, 2003. The following is being submitted in response to the additional request for information.

1. The April 14, 2003 submittal provided a precedence (Fermi Unit 2) for their relief request. The precedence contained a request for relief from UT to enhanced VT-1 with a 100% coverage and another request for relief from UT to enhanced VT-1 with less than 100% coverage.

The staffs safety evaluation considered the coverage and sample size from the request with 100% coverage when reviewing the request of less than 100% coverage. Identify other RPV inner nozzle radii that were essentially 100% examined with enhanced VT-1 (referenced previous safety evaluation). What percent of coverage were you able to achieve during the first interval using UT?

During Hope Creek's first inspection interval the reactor pressure vessel (RPV) inner nozzle radii were essentially 100% examined using manual UT at the cost of significant personnel exposure. Hope Creek performs automated UT of the Feedwater nozzles and expects to continue doing so per the requirements of NUREG-0619.

The remainder of the RPV nozzles inner radii receives a manual UT from the reactor vessel's external shell surface and nozzle bore regions. In PSEG's April 14, 2003 request, it was indicated that dose rates for specified RPV nozzles were in the range of 200 mR/hr to 250 mR/hr with shielding in place. During Hope Creek's RFO11 (Spring 2003), several of the RPV nozzles exhibited significantly elevated dose rates. The nozzles average dose rates ranged from 250 - 300 mR (nozzle N2) to approximately 9.0 R (nozzle N17) on contact. The manual UT exams are performed from within each nozzle's door opening and contribute to higher personnel radiation exposures due to the examiner's close proximity to the 1

Document Control Desk LR-N04-0002 shielded piping. Experience during RFO1 1 has shown that performance of these exams results in the receipt of an unnecessary additional 100 mR to 2.5 R of radiation exposure to NDE exam personnel.

Performance of a remotely operated enhanced VT-1 examination will be conducted in conjunction with the in vessel visual examinations.

Performance of the rector vessel nozzle inner radii will significantly reduce unnecessary additional exposure to NDE exam personnel and still maintain an adequate level of quality and safety. If the enhanced VT-1 examination detects a discontinuity requiring additional evaluation, then a supplemental manual UT exam may be performed, as needed.

2. In the September 3, 2003 response, Question 2 describes the restrictions as the thermal sleeves covering essentially 360 degrees inside the nozzle bore segment of M-N in Figure 1. If the thermal sleeve is not the sole source of the restriction, describe or provide a sketch showing the restrictions and identify the nozzle location affected by the restriction.

A typical nozzle configuration sketch for the RPV1-N8A and RPV1-N8B is shown below. This sketch identifies the twelve (12) jet pump instrumentation tubes that are encased within the N8A and N8B reactor vessel nozzles.

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Document Control Desk LR-N04-0002 The sketch below demonstrates the N8A and N8B typical reactor vessel jet pump instrumentation nozzle design that retains the twelve (12) jet pump instrument tubes as shown emanating from the nozzle as shown.

3. In the September 3, 2003 response, Question 3 describes a demonstration process of an automated visual examination system but does not address direct visual examinations. Explain the qualification process for direct visual examinations that will be used to demonstrate the 1-mil width sensitivity. Include a discussion on the application of the qualified process.

On page 1 of 4 of relief request HC-RR-B12, the statement: "Reactor vessel closure head vent and spray nozzles inner radii will receive direct visual examinations (VT-1) conducted in accordance with ASME Section XI requirements, while the other remaining aforementioned components will receive enhanced visual examinations using the 1-mil wire diameter wire standard", was an inadvertent error. Direct visual examinations are only intended to be performed upon the reactor vessel closure head's spray and vent penetrations as stated in relief request HC-RR-B11 that was previously approved by the NRC on June 9, 2003 (TAC No. MB8408).

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