ML041600502

6/8/04, Safety Evaluation for Fort Calhoun Station, Third 10 - Year Inservice Inspection Interval, Request for Relief (RR) 9

ML041600502
Person / Time
Site:	Fort Calhoun
Issue date:	06/08/2004
From:	Stephen Dembek NRC/NRR/DLPM/LPD4
To:	Ridenoure R Omaha Public Power District
Dembek S, NRR/DLPM,415-1455
References
TAC MC1115
Download: ML041600502 (8)

Text

June 8, 2004 Mr. R. T. Ridenoure Division Manager - Nuclear Operations Omaha Public Power District Fort Calhoun Station, FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

SAFETY EVALUATION FOR FORT CALHOUN STATION, THIRD 10-YEAR INSERVICE INSPECTION INTERVAL, REQUEST FOR RELIEF (RR) 9 (TAC NO. MC1115)

Dear Mr. Ridenoure:

By letter dated October 24, 2003, as revised in its entirety by letter dated November 21, 2003, Omaha Public Power District (OPPD) submitted RR-9 for the third 10-year inservice inspection interval at the Fort Calhoun Station. In response to an NRC request for additional information dated March 12, 2004, OPPD submitted a letter dated April 2, 2004, which revised the November 21, 2003, letter to request the relief under 10 CFR 50.55a(a)(3)(ii) instead of 10 CFR 50.55a(a)(3)(i). The staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the subject welds. Therefore, compliance with the ASME Code examination requirement would result in a hardship or unusual difficulty without a compensating increase in quality or safety. Pursuant to 10 CFR 50.55a(a)(3)(ii), the staff authorizes RR-9 for the third 10-year inservice inspection interval which ended October 31, 2003. The staffs safety evaluation request is provided in the enclosure.

If you have any questions, please contact Alan Wang, Project Manager, at (301) 415-1445.

Sincerely,

/RA/

Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

Safety Evaluation cc w/encl: See next page

ML041600502

• Memo dated NRR-028 OFFICE PDIV-2/PM PDIV-2/LA EMCB

• OGC PDIV-2/SC NAME AWang:mp EPeyton TChan MBupp SDembek DATE 5/19/04 5/19/04 4/26/04 6/4/04 6/8/04 DOCUMENT NAME: G:\PDIV-2\FortCalhoun\mc1115se.wpd Ft. Calhoun Station, Unit 1 cc:

Winston & Strawn Mr. Daniel K. McGhee ATTN: James R. Curtiss, Esq. Bureau of Radiological Health 1400 L Street, N.W. Iowa Department of Public Health Washington, DC 20005-3502 401 SW 7th Street, Suite D Des Moines, IA 50309 Chairman Washington County Board of Supervisors Mr. Richard P. Clemens P.O. Box 466 Division Manager - Nuclear Assessments Blair, NE 68008 Omaha Public Power District Fort Calhoun Station Mr. John Kramer, Resident Inspector P.O. Box 550 U.S. Nuclear Regulatory Commission Fort Calhoun, NE 68023-0550 P.O. Box 310 Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Sue Semerera, Section Administrator Nebraska Health and Human Services Systems Division of Public Health Assurance Consumer Services Section 301 Cententiall Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. David J. Bannister, Manager Fort Calhoun Station Omaha Public Power District Fort Calhoun Station FC-1-1 Plant P.O. Box 550 Fort Calhoun, NE 68023-0550 Mr. John B. Herman Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF RR-9 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NO. 50-285

1.0 INTRODUCTION

The inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1, Class 2, and Class 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of Record for the Fort Calhoun Station for the third 10-year ISI interval, which began on September 26, 1993, and ended October 31, 2003, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code, with no addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

By letter dated October 24, 2003, as revised in its entirety by letter dated November 21, 2003, Omaha Public Power District (the licensee) submitted relief request (RR) 9 for the third 10-year ISI interval at the Fort Calhoun Station, Unit 1 (FCS). In response to an NRC request for additional information dated March 12, 2004, the licensee submitted a letter dated April 2, 2004, which revised the November 21, 2003, letter to request the relief under 10 CFR 50.55a(a)(3)(ii) instead of 10 CFR 50.55a(a)(3)(i). In RR-9, the licensee proposed replacing the Code-required surface examination with an ultrasonic testing (UT) examination from the inside diameter (ID) surface.

2.0 DISCUSSION 2.1 Components for which Relief is Requested Reactor pressure vessel (RPV) nozzle-to-safe end, dissimilar metal welds MRC-1/01, MRC-1/18, MRC-1/30, MRC-2/01, MRC-2/18, and MRC-2/30.

2.2 Code Requirements 1989 Edition, no Addenda,Section XI of the ASME Code, Table IWB-2500-1, Examination Category B-F, Item Number B5.10, Reactor Vessel, "NPS 4 or larger nozzle-to-safe end butt welds" require a volumetric and surface examination defined by Figure IWB-2500-8.

2.3 Proposed Alternative The licensee proposed implementing an ultrasonic examination method from the ID surface for the surface examination in lieu of the Code-required surface examination for the six (6) subject RPV nozzle-to-safe end, dissimilar metal welds.

2.4 Licensees Basis for the Alternative The licensees proposed alternative is to use a UT examination method from the ID surface to examine the surface in lieu of the Code-required surface examination method. The proposed UT examination as submitted in the attachments to the letter dated November 21, 2003, is described in the non-proprietary version of the qualification documentation (Framatome ANP "Results from ID & OD [outside diameter] Clad Safe-end Mockup Block Demonstration for Fort Calhoun," 54-PQ-189-01) and the procedure for the performance of the UT examination technique used at FCS during the 2003 refueling outage to perform the surface examination of the RPV Examination Category B-F welds is described in the Framatome ANP Nondestructive Examination Procedure, "ID Automated Ultrasonic Examination of Welds for detection of OD Initiated Flaws," 54-ISI-189-01.

The UT examination techniques utilized for this examination were qualified by demonstration at the Electric Power Research Institute Nondestructive Examination (NDE) Center, Charlotte, North Carolina. The use of these qualified techniques assured that the dissimilar metal welds remain free of service related flaws thus enhancing quality and ensuring plant safety and reliability.

The surface inspections of the outside weld surfaces once accessed, are limited due to the confined space and due to the close proximity of the wall of the sand box to the outside of the pipe/nozzle. Only 60 percent of the required weld can be inspected from the OD surface, whereas 100 percent of the Code-required weld surface was inspected using the alternative UT technique performed from the ID surface.

The area radiation dose rate is estimated as being 120 mr/hr with the RPV head on the vessel.

The dose rate in the small cavity surrounding each nozzle is unknown. An ex-core detector was removed from one of the nozzle boxes last refueling outage and read 40,000 mr/hr on contact.

The surface dose rate near the welds in question would be very close to these detectors.

It is estimated that the total dose for this examination from the OD surface would be 3-6 man-rem. There is no additional dose for performing these examinations with UT from the ID surface, since all the equipment is already in place for other RPV UT examinations. Therefore, the implementation of this alternative method reduces the radiation exposure by 3-6 man-rem while providing an acceptable level of quality and safety.

The ID surface weld profilometry was performed on the RPV nozzle welds during the Fall 2003 refueling outage and no counterbore or ID profiles were detected that interfered with transducer contact during the UT examinations in either the circumferential or axial directions. The alternative examinations were performed during the Fall 2003 refueling outage and no OD surface indications were identified.

2.5 Evaluation The licensee initially requested relief for these subject welds in a letter dated December 20, 2002. After discussions with the staff, the licensee withdrew the December 20, 2002, request.

On October 22, 2003, the licensee submitted the current proposed alternative for the same subject welds.

The RPV nozzles are encased in a sandbox that must be removed to provide access to the surface of the RPV nozzles. Once the outside weld surfaces are accessed, the surface inspections are limited due to the confined space and limited access due to the RPV and piping configuration. With the sandboxes removed, the licensee stated that only 60 percent of the weld surface is inspectable with the Code-required surface examinations. The licensee proposed to examine the outside surface of the subject butt welds with UT examinations of the near surface volume in lieu of the Code-required surface examinations. The UT examination does not require direct contact or near contact with the inspection surface, thus minimizing exposure at these high radiation areas. The licensee estimates that removing the sandboxes and performing the surface examinations would result in hardship or unusual difficulty because it would result in an estimated increase of 3 to 6 man-REM exposure to the NDE personnel.

Therefore, the implementation of this alternative method reduces the radiation exposure by 3-6 man-rem.

Both examination methods are capable of detecting surface-breaking flaws but 100 percent of the surface (volume just below the surface) is inspectable with the UT examinations. However, if a surface examination detected any flaws, a follow-up UT examination would be necessary to

size the flaws. The 1989 Edition of ASME Code does not contain UT examination criteria for performing examinations from the ID surface to detect flaws on the OD surface of piping.

The subject nozzle-to-safe end welds are a hybrid corrosion resistant cladded weldment. The ferritic nozzle bore is cladded with low carbon stainless steel which is joined to a stainless steel safe end with an Alloy 182 (Inconel) weld material. The nozzle-to-safe end was heat treated as part of the RPV assembly. After heat treatment, the safe-end ID and OD surfaces were cladded with a low carbon stainless steel filler material. After cladding, the weldment ID was machined smooth and the OD was ground smooth. A profilometry examination performed on the ID weld region during the last outage verified the smooth bore, i.e., no obstructions or interferences such as misalignments or convex/concave weld roots. The smooth ID surface provided excellent conditions for UT examinations. However, the dissimilar weld and austenitic weld interfaces can present difficulties in interpreting acoustic responses and demonstration reliability.

Because of the smooth bore condition, the problems associated with transducer contact are alleviated. The inherent microstructure anomalies of austenitic material can still affect UT examination capabilities. To demonstrate the inspectability of the subject nozzle-to-safe ends, the licensee secured a mock-up similar to their nozzle-to-safe end weld configuration and materials. The mock-up had an axial crack and circumferential crack on the safe end OD that extended 0.8 percent through-wall. The butt weld had side drilled holes located at various through-wall depths in the Inconel weld material. The licensee examined the mock-up using the UT procedure and personnel for the proposed alternative examination and successfully detected and sized the cracks. The procedure specified specific transducers for the different scanning directions and focal lengths (through-wall depths). Although there probably is variability in the microstructure between the mock-up and subject welds, the successful UT demonstration with the mock-up should be reflective of the same capabilities on the subject welds.

In addition to the UT examination, the licensee performed an evaluation of the likelihood for developing OD surface cracks at the subject welds. In the letter dated December 20, 2002, the licensee stated, "These six (6) welds have no history of being susceptible to any probable leak paths which would cause external chloride stress corrosion cracking or other outside surface initiated mechanisms for outside surface cracking." In the absence of a corrosion source or mechanical source, the development of a crack from plant operations is minimal.

In summary, the existence of a detrimental OD surface crack occurring is remote, and in the event that an OD surface crack did occur, the improved coverage and flaw detectability of the UT examination should detect it, thus providing reasonable assurance of structural integrity.

With the above monitoring for surface flaws, the UT examination has the advantage of providing a better working environment by lowering the radiation exposure of the NDE personnel. Therefore, the UT examination will provide an acceptable alternative to a Code-required surface examination to which compliance would result in hardship to the licensee without a compensating increase in the level of quality and safety.

3.0 CONCLUSION

Based on the above evaluation, the staff concludes that the surface examination would result in hardship or unusual difficulty as it requires working in a high radiation area, resulting in a 3 to 6 man-rem exposure to NDE personnel. The staff has also concluded that the proposed alternative provides reasonable assurance of structural integrity of the subject welds.

Therefore, compliance with the ASME Code examination requirement would result in a hardship or unusual difficulty without a compensating increase in quality or safety. Pursuant to 10 CFR 50.55a(a)(3)(ii), the staff authorizes RR-9 for the third 10-year ISI interval which ended October 31, 2003.

All other ASME Code requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: D. Naujock Date: June 8, 2004