ML032530391

Request for Additional Information Response - Relief Request HC-RR-B12

ML032530391
Person / Time
Site:	Hope Creek
Issue date:	09/03/2003
From:	Salamon G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HC-RR-B12, LRN-03-0375, TAC MB8407
Download: ML032530391 (9)

Text

PSEG Nue LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038.0236 SEP 3 2003 LRN-03-0375 PSEG Nuclear LLC United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST HC-RR-B12 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE DPF-57 DOCKET NOS. 50-354

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 volumetric examination required by the American Society of Mechanical Engineers 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 staff discussed the subject relief request with PSEG staff on June 18, 2003, and requested additional information be provided in response to their June 26, 2003 letter. Pursuant to that request, PSEG is submitting the enclosed response to the request for additional information.

Should you have any additional questions, please contact Mr. Howard Berrick at 856-339-1862.

Sincerely, alamon Manager - Nuclear Safety and Licensing Enclosure 14 95-2168 REV. 7/99

- F Document Control Desk LRN-03-0375 C Mr. H. Miller, Regional Administrator - Region I U. S. Nuclear Regilatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. Nuclear Regulatory Commission ATTN: Mr. R. Ennis Licensing Project Manager - Hope Creek Mail Stop 08B1 Washington DC 20555-001 USNRC Senior Resident Inspector - Hope Creek (X24)

(w/o enclosure)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625 (w/o enclosure) 2

Document Control Desk Enclosure LRN-03-0375 REQUEST FOR ADDIT;ONAL INFORMATION HOPE CREEK GENERATING STATION RELIEF REQUEST HC-RR-B12 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).

Specifically, the PSEG relief request proposes to perform an enhanced remote visual examination (EVT-1) technique of the surface M-N shown in ASME Section Xl Figures IWB-2500-7 (a) through (d) as an alternative to ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100 requiring volumetric examination (Ultrasonic, UT) of the Inner Radius of Class 1 Reactor Pressure Vessel (RPV) Nozzles.

The enhanced remote visual examination will be performed upon the examination surface M-N to achieve essentially 40-60% coverage using 8x magnification video equipment to examine the inner radii. The resolution sensitivity for this remote examination will be established using a 1-mil diameter wire standard, similar to that used for other reactor pressure vessel internal examinations intended to detect cracking.

The NRC staff, in reviewing the submittal, has determined that the following information will be needed to complete the review:

S Document Control Desk Enclosure LRN-03-0375 NRC Inquiry #1:

Table I In the submittal includes the summary iumber and component identification of each nozzle to be examined. The type of nozzle to be examined is also important because nozzles subjected to large thermal gradients have a past history of thermal cracking. Describe the type (e.g.,

feedwater, recirculation inlet, jet pump instrumentation, etc.), of nominal Inside diameter and material (e.g., carbon, nickel, etc.) of the nozzles Included In the relief request.

PSEG Nuclear Response:

Table 1 below has been revised to inclUde the requested information. The table now includes descriptions for the nozzle, nozzle dimensions and material composition.

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I Document Control Desk Enclosure LRN-03-0375 Table I Hope Creek RPV Nozzle Inner Radlus Exams Summary Component Estmated Shell Inner Bore Outer Inside Outside Number identIfication ces a Limitation Confluration Material Thlickness Redus Blend Bore Bore I~ ~ ~~0e8 Radius Dimension Dimension 100195 RPV1-N2A WX% Theral Sleeve/Jet Pump Riser CA508 6 1.772 3.346 11.496' 13.975 100200 Ri'V1-N28 50% 300 Recirculation Inlet Nozzle Class 2 100200 RPV1-N2B 50% Thermal Sleeve/Jet Pump Riser A508 6. 1.772" 3.346" 11.4W t3.975' 60" Recirculatlon Inlet Nozzle Class 2 100205 RPV1-N2C 50% Thermal Sleeve/Jet Pump Riser SA508 6.7 1.72" 3.346" 11.496 13.975' 90" Recirculation Inlet Nozzle Class 2 100210 RPV1-N2D Thermal Sleeve/JetInlet 120"1 Recircuiation Pump Riser SA50 2 62. 1.772 3.346' 11.496" 13.975' Nozzle Class 4

100215 RPVI-N2E Them al Sleeve/Jet Inlet 150. Recirculation Pump Riser Nozzle CaM22 Class 6.r 1.772' 3.346' 11.496 13.975" 100220 RPV1-N2F 50% Thermal Sleeve/Jet Pump Riser SA508 67r 1.77? 3.343" 11.496' 13.975' 210"1 Recirculation Inlet Nozzle Class 2 100230 RPV1-N2H 50% l~~~~heffnal S ev~rt Pump Riser Ca08r 1m3.4i 149 395 100225 RPVI-N20 50% ~~~Thermal Sleeve/Jet Pump Riser SA50 7 .72 .48 .9 100225 RPV1-NSA 40% 240 RecSrcuaton Inlet Nozzle Class 2 8.5 1.77 3.071 9.370 11.96 10030 RPV1-N2H 40% Thermal Slee/Jet Pump Riser SA508 6." 1.772 3.349 11.495" t3.975" 270"1 Recirculation Inlet Nozzle Class 2 100235 RPVI-N2J 5 Thermnal Sleeve/Jet Pump Riser 300" Recirculation Inlet Nozzle SA508 Class 2 ex7 1.772" 3.346' 11.496 13.975' 100240 RPVI-N2K 50% Thermal Sleeve/Jet Pump Riser 330" Recirculation Inlet Nozzle SASOB class 2 6 7r 1.772" 3.346' 11.496 13.975" 100295 RPV1-N5A 100295 RPVI-N5A 40%

40%1200

~~~~~Thermal CoreSleeve Spray and InletSparger Nozzle SA508OB2 Class 1.772? 3.071" ____

9.370" 11.496" 100300 RPVI-N5B 40% Thermnal Sleeve and Sparger SA508 6.1 1.77?' 3.071" 9.370" 11.496" 240" Core Spray Inlet Nozzle Class 2 3

I I Document Control Desk Enclosure LRN-03-0375 Table I (cont'd)

Hope Creek RPV Nozzle Inner Radius Exams Summay Idenifflon Summary Estimated EstimahdComponentLimitation Configuration material shell Thle Inner Bore isRadlus ~~~OuterInside Blend Born Outside Bore Number coverage Identification Thickness Radius ~~~~~~~ ~~Radius Dimension Dimension 100320 100320 RPV1-N8A RPV1-N8A 60%

G0% ~~~~~~Instrumentation 112.50- Jet Pump nstrumentationLines Nozzle SA50B2 Class 65" 8 1.377" .98 1.968 3.819' 5.197r 100325 RPVI-N8B 60% InstumpIntation Lines SA508 6.85 1.377' 1.98 3.819" 5.19r 0% 292.5"1- Jet Pump Instrumentation Nozzle Class 2 100400 RPVI-N17A 50% Thermal Sleeve/ Collar & Bolt Assembly SAss 6.7 1.7720 3.386 11.378 - 13.976 450 LPCI Inlet Nozzle Class 2 100401 RPVI-NI7B 50% Thermal Sleeve/ Collar & Bolt Assembly SA508 6.7' 1.772" 3.386' 11.378 13.976' 135", LPCI Inlet Nozzle Class 2 Thermal Sleeve! Collar &Bolt Assnmhly SA508 11- . -, "I 100403 RPV1N17D 50% 6.7' 1.772 11.378 13.976' 315 LPCI Inlet Nozzle Class 2

,, 1......

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Document Control Desk Enclosure LRN-03-0375 NRC Inquiry #2:

The licensee states the estimated coverage of each nozzle In Table I of Its submittal. The list fails to state specific information about which area of the nozzle will be inaccessible to examinetion (e.g. top, side, bottom). The severity of thermal fatigue is dependent upon the circumferential location, thermal gradient and thermal sleeve leakage. Describe the portion of each nozzle that will be Inaccessible to examination, providing a sketch If possible.

PSEG Nuclear Response:

In the original request PSEG stated that certain Hope Creek RPV Nozzle Inner radius exams contained configurations that impeded complete 100 percent visual examination coverage of the nozzle inner radius area surface M-N Hope Creek's Low Pressure Core Injection, Core Spray and Recirculation Systems inlet nozzles, and Jet Pump Instrumentation nozzles.

The enhanced remote visual examination will be performed upon the examination surface M-N to achieve essentially 40-60% coverage using 8x magnification video equipment to examine the inner radii. The resolution sensitivity for this remote examination will be established using a 1-mil diameter wire standard, similar to that used for other reactor pressure vessel internal examinations intended to detect cracking.

The area of the nozzle that is inaccessible is portrayed in Figure 1 below and should be considered to be 3600 around the circumference of the nozzle inner radii area and limited to the approximate surface area shown within between M-N.

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Document Control Desk Enclosur6 LRN-03-0375 I'

tn 1 _

K I 0

P TheralSee_

Thermal Sleeve be examined

• T-Cladding interior where present Figure 1 6

Document Control Desk Enclosure LRN-03-0375 NRd Inquiry #3:

In the alternative, the licensee states that the ri.-zliuton sensitivity will be established using a 1-mil diameter wire standu that is similar to that used for other reactor pressure vessel nternal examinations intended to detect cracking without stating which standards are being referenced. Explain the qualification process that will be used to demonstrate te 1-.Al width sensitivity and the equipment used for the examination.

PSEG Nuclear Response:

The qualification process used to demonstrate the 1-mil width sensitivity is as follows:

PSEG 1 will fabricate a Sensitivity, Resolution and Contrast Standard (SRCS) that is representative of the surface texture (reiectivity, color and finish) of the item to be examined. Targets of sufficient length to demonstrate the required resolution across the entire field of view of the camera system are affixed or embedded into the SRCS. At least one such target will be oriented in the horizontal direction and another target oriented in the vertical direction. The-target is a wire of less than or equal to 1- mil width.

Equipment resolution and sensitivity is demonstrated prior to performing examinations.

Resolution and sensitivity of the examination equipment and technique is considered adequate when the system is capable of discerning the required target (that is, the 1-mil diameter wire standard).

The requirements for the video equipment used in the examination will be consistent with BWRVIP-03, Revision 3, Reactor Pressure Vessel and Intemals Examination Guidelines, Section 2.5 Generic Standards for Visual Inspection of Reactor Pressure Vessel Internals. Components and Associated Repairs. Parameters considered in the nozzle inner radii visual exams include lighting, depth of field, field of view, magnification, and speed of camera movement.

' PSEG Nuclear will contract an !SI vendor to perform the ISI inspections.

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