ML20059D074
| ML20059D074 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/29/1993 |
| From: | Siegal B Office of Nuclear Reactor Regulation |
| To: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| References | |
| TAC-87122, TAC-87123, NUDOCS 9401060398 | |
| Download: ML20059D074 (6) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 29, 1993 Docket Nos. 50-348 and 50-364 Mr. D. N. Morey, Vice President Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
Dear Mr. Morey:
SUBJECT:
RELIEF REQUESTS 41 AND 42 FOR SNUBBER TESTING FOR SECOND l
TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM FOR ASME CODE CLASS 1, 2 AND 3 COMPONENTS - JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 & 2 (TAC NOS. M87122 AND M87123)
By letter dated November 23, 1987, you submitted Revision 0 of the subject Inservice Inspection (ISI) Program for staff review.
It was recently brought to our attention that Relief Requests (RRs) 41 and 42 which are related to snubber testing have never received staff review and, therefore, have remained open in the subsequent revisions of the program.
The staff has since reviewed these relief requests, as. listed in Revision 0 of the ISI orogram, and find them acceptable.
Our evaluation, which is applicable to both units, is.provided below.
Relief Reauest 41 - Break-Away Draa Test for Hydraulic Snubbers Relief was requested from the test requirement for measuring break-away force during low velocity displacements, as required by Paragraph IWF-5400(b)(1) of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, 1983 Edition, for hydraulic snubbers. The licensee stated that hydraulic snubbers are inherently inconsistent when this particular test is considered. The load required to initiate movement will vary considerably between identical snubbers depending on a number of variables, such 'as viscosity of hydraulic fluid, valve spring rate, temperature, etc.
i Because of the number of variables that affect the break-away force, the manufacturers of hydraulic snubbers have not recognized break-away force as a test variable that is relevant to snubber integrity and this parameter is not addressed in their technical information.
The staff has determined that this test requirement is impractical based on the fact that break-away force is not relevant to snubber integrity due to the number of variables that affect this parameter.
The staff.also acknowledges-1 the current industry trends in excluding the measurement of break-away force as a portion of the requirements of the. snubber functional testing.
In fact, a possibility exists that such a requirement will be taken o'ut from the future edition of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). -This potential modification of code requirements is a result I Di of an improved industry understanding over the past years in the area 9401060398 931229 PDR ADOCK 05000348 Ll 050028 E ME CEMER C&PT
- of snubber behavior. The staff.would, therefore, considers it a burden of compliance if the licensee is required to test snubbers based on. criteria which have already been proved to be impractical and meaningless. - Pursuant to 10 CFR 50.55a(g)(6)(1), the Commission may grant relief and may impose.such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense.and security and is otherwise in the public interest giving d1e consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
However the granting of this relief is predicated on your having an adequate basis for the acceptance criteria for snubber functional testing, in the L
Joseph M. Farley Nuclaar Plant, Units 1 and 2 (Farley), Technical Specifications (TC).
Relief Reouest 42 - Additional Samole Testino Reouirements For Snubbers Paragraph'IWF-5400(c) of ASME Code,Section XI, 1983 Edition, requires an_
additional sampling of 10 percent of the total number of snubbers to be tested if the initial 10 percent representative sample has failed inservice testing.
Additional sample testing is to continue until all snubbers within the sample have passed the test.
1 The licensee stated that the above additional sampling plan of IWF-5400(c) is more restrictive than the existing expansion plan detailed in TS 3/4.7.9c, which requires an additional sampling plan of 35(1+C/2), where C is the number of snubbers not meeting the acceptance criteria. The basis of the above expansion plan is' that selection of such representative additional samples provides a confidence level of approximately 95 percent that 90 to 100 percent of the snubbers in the plant will be operable within acceptance limits.
The staff's evaluation revealed that the Section XI sample plan would require more testing than "The 10% Testing Sample Plan" (10% plan) recommended by the ASME OM Code, 1990 Edition.
According to the ASME OM Code, an additional sample of at least one-half the size of the initial sample (or 5%, instead of 10% as required by Section XI) shall be tested until the total number tested is equal to the initial sample size multiplied by the factor 1+C/2, where C is the total number of the snubbers found to be unacceptable.
The staff.also finds that the expansion plan specified by the licensee in their TS is comparable to "The 37 Testing Sample Plan" (37 plan) which, like the 10% plan, is recommended by the ASME OM Code and is an acceptable alternative to the staff. Since implementation of either the Section XI sample plan or the ASME OM Code "37 plan" is acceptable to the staff, to the Section XI sample plan need not be implemented if the '37 plan" or its equivalent, the TS expansion 1
t t plan, is implemented. Since the proposed alternative provides an acceptable level of quality and safety, pursuant to 10 CFR 50.55a(a)(3)(1) this request far relief from the Section XI requirement is granted.
Sincerely, Byr L. Siegel, ting Project Director Pr ect Director e 11-1 Division of Reactor Projects - I/II Office of Nuclea'. Reactor Regulation ec: See Next Page 4
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c December 29, 1993 l plan is implemented.
Since the proposed alternative provides an acceptable l
level of quality and safety, pursuant to 10 CFR 50.55a(a)(3)(i) this request for relief from the Section XI requirement is granted.
Sincerely, p
1 N'A V
By n L. Siegel, cting Project Director Project Directorate 11-1 i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation i
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- See previous concurrence OFC LA9Qi(hkPE PM:PD21f0)(PE PD:PD21aD9PE-OGC*
NAME PAnkon BSidkjrm SBajwak MYoung DATE 12/43/93 12/[i/93 12/ # /93 12/16/93 0FFICIAL RECORD COPY DOCUMENT NAME: G:\\FARLEY\\FAR87122.REL l
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.h Mr. D. H. Morey Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc.
CC; l
Mr. R. D. Hill, Jr.
State Health Officer General Manager - Farley Nuclear Plant Alabama Department of Public Health Southern Nuclear Operating Co., Inc.
434 Monroe Street Post Office Box 470 Montgomery, Alabama 36130-1701 Ashford, Alabama 36312 Chairman Mr. B. L. Moore, Licensing Manager Houston County Commission Southern Nuclear Operating Co., Inc.
Post Office Box 6406 Post Office Box 1295 Dothan, Alabama 36302 Birmingham, Alabama 35201-1295 Regional Administrator, Region II James H. Miller, III, Esquire U. S. Nuclear Regulatory Commission Balch and Bingham Law Firm 101 Marietta St., N.W., Ste. 2900 Post Office Box 306 Atlanta, Georgia 30323 1710 Sixth Avenue North Birmingham, Alabama 35201 Resident Inspector _
U.S. Nuclear Regulatory Commission Mr. J. D. Woodard Post Office Box 24 - Route 2 Executive Vice President Columbia, Alabama 36319 Southern Nuclear Operating Company P.O. Box 1295 i
Birmingham, Alabama 35201 i
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LDISTRIBUTION
%EDocket? File" NRC & Local'PDRs PD#I1-1 Reading T. Murley/F. Miraglia L. J. Callan, Acting E. Rossi J. Lieberman S.'Varga G. Lainas-S. Bajwa P. Anderson B. Siegel OGC E. Jordan G. Hill ACRS (10)
OPA OC/LFDCB L. Plisco, E00 E. Merschoff, Region II cc:
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