ULNRC-05426, Day Report for First Revised NRC Order EA-03-009 (Regarding Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors)

Day Report for First Revised NRC Order EA-03-009 (Regarding Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors)

ML071980370
Person / Time
Site:	Callaway
Issue date:	07/09/2007
From:	Graessle L AmerenUE
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EA-03-009, ULNRC-05426
Download: ML071980370 (7)

Text

AmerenUE PO Box 620 CallawayPlant Fulton, MO 65251 July 9, 2007 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Ladies and Gentlemen: ULNRC-05426 DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

WAmeren 60-DAY REPORT FOR FIRST REVISED NRC ORDER EA-03-009 (REGARDING INTERIM INSPECTION REQUIREMENTS iE FOR REACTOR PRESSURE VESSEL HEADS AT PRESSURIZED WATER REACTORS)

Reference:

AmerenUE letter ULNRC-04957, "Response to Order EA 009, Revision 1, 'Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors,"' dated March 5, 2004 On February 11, 2003 the NRC issued Order EA-03-009 providing interim inspection requirements for reactor pressure vessel (RPV) heads at pressurized water reactor (PWR) facilities. The order was subsequently revised, and on February 20, 2004 the NRC issued First Revised Order EA-03-009 (i.e., "First Revised Order Modifying Licenses") thus superseding the original Order. In response to First Revised Order EA-03-009, Union Electric Co. (AmerenUE) submitted a letter dated March 5, 2004 (above Reference) wherein AmerenUE consented to the revised NRC Order for Callaway Plant.

Sections IV.C and IV.D of the revised Order specify requirements for the performance of inspections of the RPV head and head penetration nozzles, and for inspections to identify potential boric acid leaks from pressure-retaining components located above the RPV head, respectively. These inspections are to be performed during refueling outages in accordance with the schedule specified in the revised Order.Section IV.E of the revised Order requires that for each examination required in Section IV.C, the Licensee shall submit a report detailing the examination results within 60 days after returning the plant to operation.

Similarly,Section IV.E also requires that for each examination required in Section IV.D of the revised Order, the Licensee shall submit a report detailing the examination results within 60 days after returning the plant to operation if a leak or boron deposit was found during the examination.

4to(

a subsidiaryofAmeren Corporation

ULNRC-05426 July 9, 2007 Page 2 of 3 Pursuant to the report requirements of the revised Order, AmerenUE hereby reports that during Callaway's fifteenth refueling outage (RF 15), which was concluded on May 10, 2007, a visual examination to identify potential boric acid leaks from components located above the reactor pressure vessel (RPV) head was performed. No leakage or boron deposits were observed. Further, non-visual nondestructive volumetric examinations of all 78 control rod drive mechanism (CRDM) penetrations, as well as the head vent penetration, were performed. From these examinations, no indications of cracks or leakage pathways were identified, and no flaws needing disposition or corrective action were identified.

Bare metal visual examinations of 100 percent of the RPV head penetrations were also performed. Included was a 360-degree examination around each of the vessel head penetration nozzles as well as the head vent penetration.

Visual examination of the RPV head surface was also performed to identify any degradation. From these examinations, no evidence of active leakage or degradation of the RPV head was identified.

Additional details regarding the above-described examinations/inspections are provided in Attachment 1.

No new or additional regulatory commitments are made in this letter. For any questions regarding this report, please contact me at 573-676-8129 or Scott A.

Maglio at 573-676-8719.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on: J, (q1 'i 0o -7 Luke H. Graessle Manager - Regulatory Affairs JAD/TBE/slk

Attachment:

60-Day Report for First Revised NRC Order EA-03-009

i ULNRC-05426 July 9, 2007 Page 3 cc: Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-7D 1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street PO Box 360 Jefferson City, MO 65102-0360

ULNRC-05426 July 9, 2007 Page 4 bcc: C. D. Naslund A. C. Heflin L. H. Graessle K. A. Mills G. A. Hughes S. A. Maglio S. L. Gallagher (100)

L. M. Belsky (NSRB)

T. A. Moser D. M. Stepanovic J. A. Doughty A160.0761 Certrec Corporation 4200 South Hulen, Suite 422 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed).

Send the following without attachments:

Ms. Diane M. Hooper Mr. Dennis Buschbaum Supervisor, Licensing TXU Power WCNOC Comanche Peak SES P.O. Box 411 P.O. Box 1002 Burlington, KS 66839 Glen Rose, TX 76043 Mr. Scott Bauer Mr. Stan Ketelsen Regulatory Affairs Manager, Regulatory Services Palo Verde NGS Pacific Gas & Electric P.O. Box 52034, Mail Stop 104/5/536 Mail Station 7636 P.O. Box 56 Phoenix, AZ 85072-2034 Avila Beach, CA 93424 Mr. Scott Head Mr. John O'Neill Supervisor, Licensing Pillsbury Winthrop Shaw Pittman LLP South Texas Project NOC 2300 N. Street N.W.

Mail Code N5014 Washington, DC 20037 P.O. Box 289 Wadsworth, TX 77483

k Attachment 1 to ULNRC-05426 60-Day Report For First Revised NRC Order EA-03-009 (Regarding Interim Inspection Requirements For Reactor Pressure Vessel Heads At Pressurized Water Reactors)

ULNRC-05426 July 9, 2007 Page 2 of 3 60-Day Report for First Revised NRC Order EA-03-009 References

1. First Revised Order Modifying Licenses, EA-03-009, issued via NRC letter dated February 20, 2004, "Issuance of First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors"

2. AmerenUE letter ULNRC-05416, "Request for Relaxation from Requirements of First Revised NRC Order EA-03-009 (Regarding Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors)", dated 18 May 2007.

NRC Required Information:

The first revision of NRC Order EA-03-009, dated February 20, 2004, requires in Section IV.E. that, for each examination required in Section IV.C, the Licensee shall submit a report detailing the examination results within 60 days after returning the plant to operation, and for each examination required in Section IV.D, the Licensee shall submit a report detailing the examination results within 60 days after returning the plant to operation if a leak or boron deposit was found during the examination.

Callaway Response:

In accordance with AmerenUE's commitments to Revised Order EA-03-009 (as established in AmerenUE's response to the revised Order per Reference 1), the examinations required in Sections IV.C.(5)(a), IV.C.(5)(b), and IV.D of the revised Order were performed during Callaway's 15th refueling outage (RF-15) which was concluded on May 10, 2007. Results of the examinations are as follows.

At the beginning of the outage, a visual examination of pressure retaining components above the reactor pressure vessel (RPV) head in accordance with section IV.D of the revised Order was performed. No leaks were detected and no boron deposits were observed on the components above the RPV head or on top of the insulation.

The bare metal visual examination required in Section IV.C.(5)(a) of the revised Order was performed. No evidence of boric acid leakage or degradation was detected. The examination of the penetrations and head surface area inside the insulation was performed using a video camera delivered by a robotic crawler as well as a video probe inserted manually under the insulation. The surface area of the reactor vessel head outside the insulation down to the vessel flange was examined directly. Noted during the

i ULNRC-05426 July 9, 2007 Page 3 of 3 direct examination outside of the insulation were five minor, white stains. Three of the stains consist of residue left over from cleaning performed in RF 13. Of these, two are translucent water drop trails and the remaining one is a white "scuffed" area caused by a scratch pad from the cleaning of residue from a spill. One of the other stains is a small white dot about 1/4" in diameter with no thickness, and the other stain appears to be white dye penetrant developer solution smeared on the side of the head lifting rig. None of these stains is associated with active leakage or vessel degradation.

Non-visual non-destructive examination (NDE) of all 78 control rod drive mechanism (CRDM) penetration tubes and the one vent line penetration tube was performed in accordance with section IV.C.(5)(b) of the revised Order. The CRDM penetration tubes were subject to ultrasonic testing (UT) from the inside diameter. No crack-like indications or UT leak path signals were detected. Examination of the vent line penetration was also performed ultrasonically, along with a dye-penetrant examination of the bottom of the penetration and the J-groove weld. No crack-like indications were detected.

It should be noted that in accordance with section IV, paragraph F of the revised Order, and by letter dated May 18, 2007 (Reference 2), AmerenUE requested relaxation from the nondestructive testing requirements specified in section IV, paragraph C.(5)(b) for five of the RPV head penetrations nozzles, i.e., numbers 74 through 78. The request was to limit the required distance for examination below the lower-most portion of the J-groove weld to 0.39 inches which is where the stresses decay to below 20 ksi tension.

The actual coverage obtained below the lowest point at the toe of the J-groove weld for penetrations 75, 76, and 79 was greater than the one-inch minimum per the Order, but the coverage for penetrations 74 and 77 was limited to 0.67 and 0.83 inches, respectively.