ML20128F973
| ML20128F973 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/10/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128F946 | List: |
| References | |
| NUDOCS 9302120099 | |
| Download: ML20128F973 (3) | |
Text
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SAFETY EVALVATION BY THE OFFICE OF NOCLEAR REACTOR REGULATION RELATED TO REllEF RE0 VEST JUMBER CR-16 COMMONWEALTH EDISON COMPANY OVAD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 QOCKET NOS. 50-254 AND 50-265 1.0 LNTRODUCTION Technical Specifications (TS) for the Quad Cities Nuclear Power Station, Units-1 and 2, state that inservice inspection of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Code and applicable 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)(1).
10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used when authorized by the NRC if (i) the proposed alternatives would provide an acceptabit level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulties 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 preservice examination requirements, set forth in the ASME Code,Section XI, "kules for Inservice Inspection cf 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 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) on the date twelve months prior to the date of issuance of the operating license, subject to the limitations and modifications listed therein. 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.
Pursuant to 10 CFR 50.55a(g)(5), if a licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of a licensee's determination, the NRC staff may grant the relief in accordance with 10 CFR 50.55a(g)(6)(i) or authorized alternatives in accordance with 10 CFR 50.55a(a)(3).
9302120099 930210 DR ADOCK 0500 4
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By letter dated September 30, 1991, as supplemented on March 12, 1992, and_
August 28, 1992, Commonwealth Edison Company (CECO, the licensee) submitted a request for relief from certain ASME Code requirements for the Quad Cities Nuclear Powar Station, Units 1 and 2, during the second inspection interval.
The staff's evaluation is discussed below.
2.0 EVALUATION Relief request number CR-16 for Quad Cities Nuclear Power Station, Units' I and 2, concerned the surface examination of the attachment welds between the reactor vessel shells and the reactor vessel stabilizer brackets.
Each reactor vessel has eight stabilizers that have integral ~ attachments to their shell course No. 4.
Code Reauirement.: Section XI, IWB-2500-1, Examination Category B-H, Item B8.10 states that surface examinations (Liquid-Penetrant Examination or Magnetic-Particle Examination) are required over 100% of the welds of each integrally welded attachment to a reactor vessel.
Licensee's Code Relief Reauest: The licensee requested relief from surface examinations of attachment welds between the reactor vessel and its stabilizer brackets for Quad Cities Nuclear Power Station, Units 1 and 2.
Licensee's Proposed Alternative Examination:
The-licensee proposed that VT-1 visual examinations (to visually determine the condition of the part or component) will be performed on the accessible portions of the weld surfaces.
The licensee estimated that 100% of the top horizontal surfaces of the welds and approximately 78% of the vertical surfaces of the welds could be inspected.
Licensee's Basis for Reauestino Relief: The licensee stated.that the bottom welds of the stabilizer brackets are inaccessible for visual or surface' examination because of its proximity to the-top of'the biological. shield,. and
-to inspect the side and top attachment welds of the. stabilizer brackets, insulation must be removed to allow for a surface examination. The' licensee stated that in order to physically remove the upper and lower insulation panels at each of the eight stabilizer brackets, the lower insulation panel must be removed (and replaced) through the openings in_ the-biological shield at the~ core spray and feedwater nozzles. ' This would require removal (and reinstallation) of additional-panels at the level of the nozzles and to -
perform the required surface examination.of the.-stabilizer bracket welds, both of the adjacent upper and-lower insulation' panels must be removed.
Removal of the insulation would require the _ removal of screws holding the panels of-insulation together. The screws. holding =the insulation panels together are: located directly behind a bar running _ horizontally through the reactor stabilizer brackets. The clearance between this bar and the insulation (approximately two inches) does not allow expeditious removal _ of.
the screws. Tests indicate that the destructive removal of the screws will-
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y result:in permanen't deformation of the. screw holes of the lower insulation panel.. This deformation would require replacement of thel lower insulation panel at;each stabilizer bracket.
1 The licensee's basis for the request for relief is' impracticality regarding' the removal of insulation. panels. The licensee estimates that the initial:
cumulative exposure to' remov'e, replace, and align the. lower panels would be 112 person-rem and that the cumulative exposure (on ~a unit basis) represents 1 20% of the annual exposure goal.
Evalugt.ip_q:
The Code required surface examination of the stabilizer brackets =
is a hardship for the' licensee to perform because-extensive redesign and.
reconstruction of the shield wall would be' required and to remove, replace ~,
and align the lower panels the initial cumulated exposure would be 1121 person-rem.
It-is the staff's judgment. that the smallLincremental increase'in: planti safety that' would result from imposition of the requirement to inspect 100%'of-the attachment welds of the stabilizer brackets to the reactor vessel does not' warrant this burden. The staff determined that the licensee's proposed alternative to perform VT-1 visual examinations' on-the accessible ' portions of the weld surfaces should provide reasonable assurance of the structural:
integrity of the. reactor vessel stabilizer bracket welds.
3.0 CONCLUSION
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee hasl determined that conformance with certain Code requirements is impractical for its-facility and; submitted supporting information. The staff has concluded that the licensee' has demonstrated that the Code required 100% turface examination of-the-attachment welds of the stabilizer bracketsito the reactor vessel would result-in a hardship or unusual difficulty without a compensating increase in the' level of quality and safety.
The staff also concluded that the: 3roposed-alternative examination should= provide reasonable--assurance of tle: structural!
integrity.of the reactor vessel stabi_11zer brackets. Therefore,; pursuant to 10 CFR 50.55a(a regarding a:requ)e(3)(ii):and based on the information provided by the licensee -
st not-to >erform the Code examinations on 100% of the attachment welds on-the. sta)ilizer brackets to the reactor vessel, relief:
-request number CR-16 may be authorized.
Principal Contributor:
D. Smith T. McLellan Date: February 10, 1993 s