ML18036B261
| ML18036B261 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/27/1993 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9305040216 | |
| Download: ML18036B261 (18) | |
Text
A.CCELERATED DOCUMENT DISTRIBU'IION SYSTEM REGULAT(
INFORMATION DISTRIBUTIOSTEM (RIDS)
ACCESSION NBR:9305040216 DOC.DATE: 93/04/27 NOTARIZED: NO'OCKET FACIL:,50-2t20 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH.NAME AUTHOR AFFILIATION SALAS,P.'ennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards Request for Relief SPT-5 from ASME Section XI inservice sys pressure test program to repair flaw detected in austenitic stainless steel pipe to valve weld.
Requirement would entail pressurization of entire RPV.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR 3 ENCL t
SIIFE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME ode I
NOTES:
RECIPIENT ID CODE/NAME PD2-4 ROSS,T.
COPIES LTTR ENCL 1
0 2
2 RECIPIENT ID CODE/NAME HEBDON,F COPIES LTTR ENCL 1
1 INTERNAL: ACRS NUDOCS-ABSTRACT OGC/HDS3 RES MILLMAN,G'XTERNAL:
EGSG BROWNFB NRC PDR 6
6 NRR/DE/EMEB 1
1 OC/LFMB 1
0
~RE TZ E-Ol 1
1
~RES/DSIR/
B 1
1 EGGG RANSOME',C 1
1 NSIC 1
1 1
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1 1
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1 1
1 NOTE TO ALL'"RIDS"RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT 'fHE DOCUMEN'f CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO.ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES 'REQUIRED:
LTTR 21 ENCL 18
4 0
I'
Tennessee Valley Authority. post Office Box 2000. Decatur. Alabama 35609 APR 2V 1993 10 CFR 50.55a(g)(5)(iii)
U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Mashington, D.C.
20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No.
50-260 BROWNS FERRY NUCLEAR PLANT (BFN) AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI INSERVICE SYSTEM PRESSURE TEST PROGRAM REQUEST FOR RELIEF In accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.55a(g)(5)(iii),
BFN is submitting a Request for Relief from the specified Section XI system pressure testing requirements of the 1986 Edition of the ASME Boiler and Pressure Code for NRC review.
The details of the Request for Relief and technical basis are contained in the enclosures.
During the current Unit 2, cycle 6 refueling outage, certain replacements and repairs have been made which require hydrostatic testing.
Due to the absence of isolation valves between these areas and the reactor pressure vessel (RPV), compliance with the Section XI hydrostatic testing requirement would entail the pressurization of the entire RPV.
In view of the difficulty in performing a hydrostatic test of the RPV, TVA proposes to defer the ASME Section XI Code required hydrostatic test until the scheduled reactor vessel ten year hydrostatic test, and as an alternative, the performance of a leakage test at a pressure equal to 100 percent nominal operating pressure.
As presented in Enclosure 2,
TVA believes that the proposed alternative test provides a sufficient margin of safety.
93050402l6 930427 PDR ADOCK 05000260 P
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U.S. Nuclear Regulatory Commission APR 2V )993 TVA requests NRC approval of the Request for Relief prior to May 6, 1993.
This short reviev period is necessary in order to support the BFN
.target date for return to service from the, Unit 2, cycle 6 refueling outage.
If you have any questions, please telephone me at (205) 729-2636.
Sincerely, Pe ro alas Manager of Site 'Licensing Enclosures cc (Enclosures):
NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637
.Athens, Alabama 35611 Mr. Thierry M. Ross, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 U.S.. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
I 4l Il I
ENCLOSURE 1
REQUEST FOR RELIEF SPT-5 System:
(a) Core Spray (75),, (b) Reactor Vessel Drains 6
Vents (10)
Drawing:
Component:
Class:
2-47E817-1 (a) 12 inch,weld repair (b) 2 two-inch socket welds Function:
(a) Connects 12 inch core spray valve HCV-75-27 to 12 inch to 10 inch reducer at reactor vessel nozzle N5A (b) Reactor pressure vessel bottom head drain manual valve 2-10-505 to reactor water cleanup system Impractical Test Requirement:
IWA-4400(a) After repairs by welding on the pressure retaining boundary, a system hydrostatic test shall be performed in accordance with IWA-5000.
Basis For Relief:
The weld repair and replacement areas are situated in their respective systems between the reactor pressure vessel
'(RPV) and the first isolation valve off the vessel.
These locations do not provide a means to isolate the repaired areas from the RPV for the purpose of the hydrostatic test required by Article IWA-4000.
r The performance of a hydrostatic pressure test.of the repaired areas by pressurizing the RPV would require that 8 of the 13 main steam relief valves be removed or gagged, at an estimate associated exposure of 0.48 man-rem, and that the reactor vessel high pressure SCRAM switch be disabled.
In view of the hardship associated with pressurization of the RPV during the performance of the hydrostatic pressure, TVA requests that the performance of the required hydrostatic pressure test be deferred until the next ten year reactor vessel hydrostatic test which is scheduled near the end of the current inspection interval.
41
The integrity of the weld repairs have been ensured by the following nondestructive examinations:
a) the 12 inch core spray piping overlay repair was examined using volumetric (UT) and surface (PT) techniques; b) the two 2 inch socket welds were surface examined (PT)
In addition, each of these areas will be.subject to a system leakage test performed at nominal operating pressure (1005 psig at the RPV dome) prior to startup following each refueling outage.
Based on the Enclosure 2 evaluation provided to TVA by Structural Integrity Associates, Inc.,
San Jose, California, the decrease in the margin of safety due to the deferral of the hydrostatic test requirement is minimal.and does not present an undue risk to the plant or the public.
Alternate Testing.'he 12 inch weld overlay repair weld and the two 2-inch socket welds will receive a leakage test at full system operating pressure (1005 psig at the.RPV dome) duri'ng the Class 1 leakage test at unit startup.
A hydrostatic pressure test of the repairs will be performed in conjunction with the ten year reactor vessel hydrostatic test near the end of the current inspection interval.
~I
3150 Almaden Expressway, Suite 145 San Jose, CA 95118 (408) 978.8200 FAX:(408) 978-8964 or (408) 9784)438 ENCLOSURE 2
STRUCT KNTEajIIKTT
'3I ASSOCJATES lNC April 24, 1993 HLG-93-020, Rev. 1 Fossil Plant Operations 66 South MillerRoad Suite 206 Akron, Ohio 44333 (216) 8644886 FAX:(216) 8644705 Mr. Ed Hartwig, Tennessee Valley Authority Browns Ferry Nuclear Plant P. O. Box 2000 Decatur, AL 35602
Subject:
Technical Justification for Relief from Hydrostatic Test Requirements
Dear Ed:
During the 1993 outage at Browns Ferry Unit 2, repairs were performed on the core spray
,system (to repair a flaw detected in an austenitic stainless steel pipe to valve weld) and on the reactor vessel drain line (replacing a 2" stainless steel valve and associated stainless steel piping). Both repairs are in Class 1 equivalent locations.
The ASME Code,Section XI, which governs these repairs, requires a hydrostatic test following these repairs, since the pressure boundary in each case was'breached.
The purpose of this letter is to provide justification for requesting relief from this requirement, with an initial service:leak test to be performed now and the hydrostatic test postponed until the next regularly, scheduled hydrotest. Allother Code required NDE will also be performed now as part of repair acceptance.
1.0 HISTORICALPERSPECTIVE The hydrostatic test was originally conceived as a means for "proof testing"'f pressure retaining components prior to service.
The component was intentionally overloaded via overpressure,
.with the intent that if an undetected flaw existed which could lead to component failure.under service, the overloading would:lead to failure during the test rather than during service. If the component did not fail during the test, it was considered that service adequacy had been effectively demonstrated.
Ifthe component did fail during the test,,at least the consequences were less severe than a failure during service.
The one-time proof test approach has been an integral part of the ASME Boiler and Pressure Vessel'odes since their inception in the early 1900s.
Related pressure test requirements were included in the original ASME Section XI Code in the early 1970s without much consideration of the'limitations regarding validity of such tests for inservice
'nuclear plant components.
19S3 Celebrating l0 YeaI s ofEngineering C.rcellence 1993
41.
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Page 2 Mr. Ed Hartwig April 24, 1993 HLG-93-020, Rev. 1 2.0 LIMITINGASSUMPTIONS IN HYDROSTATICTESTS There are several assumptions which underlie the performance of hydrostatic proof tests, and which inherently limittheir validity. These assumptions are summarized in the following.
1.
Since the overload mechanism is overpressurization by a factor on design or service
- pressure, the assumption is that pressure is the governing loading mechanism.
2.
The destructive aspect of the proof test assumes that the failure mode is such that failure would occur under a one-time steady load.
This implies that the material behavior is brittle under the test conditions.
3.
Also implicitin the proof test concept is the assumption that there is no significant mechanism for flaw character to change under service conditions.
Subcritical law growth mechanisms such as fatigue, corrosion, and stress corrosion are not addressed.
However, the validity of these assumptions is clearly questionable for inservice testing of nuclear pressure vessels and piping.
In nuclear components, other loading mechanisms than pressure are generally present, and can dominate the total service loading on the component.
These loads include contributions from design seismic loading, thermal transients, thermal expansion, and dead weight, among others.
The materials typically used in nuclear components generally exhibit significant ductility. For ferritic materials, both test and service conditions are required to be well above the material brittle-to-ductile transition temperature.
For austenitic materials, such a transition is not a characteristic ofthe materials, and ductile behavior always pertains.
Consequently, although a pressure test can detect leakage, it can rarely produce an overload failure because of the ductility of the materials.
In addition, most service failures in components in nuclear systems have been traced to slow flaw growth, e.g., by fatigue or stress corrosion, rather than to the initial existence of a critical flaw. Slow growth mechanisms are not detectable by the proof testing approach.
3.0 ASME CODE ACTIONS ASME Subcommittee XI is currently re-evaluating the existing inservice pressure test requirements for nuclear components, under a special task group.
It is likely that the outcome of this review will be a revision to the Code reflecting a single "Leak-Test" requirement, in lieu of the present leak test and hydro-test requirements.
Such a change has already been implemented via Code Case N-498 for the regularly scheduled 10-year IFl1IUCWIBIRL
0
Page 3 Mr. Ed Hartwig April'24, 1993 HLG-93-020, Rev. 1 hydrotest, and similar Code action is in progress for the hydrotests associated with through-wall repairs.
4.0 HARDSHIPS IMPOSED BY PERFORMING HYDROSTATIC TESTS AT BROWNS FERRY UNIT 2 The above arguments support the position that no real additional assurance of safety is obtained by performing a hydrostatic test in addition to an initial service leak test.
Another consideration is the hardships which are imposed on the plant by requiring such a test following the repairs at Browns Ferry. In order to perform a hydrostatic test following the core spray repair, 8 out of 13 main steam relief valves would have to be gagged in order to sustain the required pressure.
The estimated total exposure associated with this task is 0.48 man-rem.
The reactor vessel high pressure SCRAM switch would also need to be bypassed.
5.0 CONCLUSION
S Although an overpressure test in the form of a hydrostatic test may be effective as an "initial service leak test" (using B31.1 terminology) following a repair or replacement, it is not significantly more effective than a leak test conducted at or slightly over operating pressure.
In fact, in the limit,of high test temperature, the required pressures for-the system leakage test.and the system hydrostatic test are nearly equal (1.0 times operating pressure for the leakage test versus 1.02 times operating pressure for the hydrostatic test,,according to Table IWB-5222-1 of Section XI). Neither test is likely to be effective in detecting non-through-wall flaws which may exist as a result of the repair or replacement (fabrication defects), nor, of course, will they detect minor flaws which will subsequently propagate by a subcritical mechanism.
Consequently, although a pressure leak'test serves a function in qualifying repairs, it is not necessary to perform two tests which perform the same function (hydrostatic and leak test),
and the hydrostatic test does not effectively perform its original function of preventing failures in. service for these ductile materials.
The costs in personnel exposure, schedule, and man-hours for performing this hydrostatic test are not justified by,the minimal additional assurance achieved by the test. Consequently, we believe that relief from the Code hydrostatic test requirement for the Browns Ferry Unit 2 repairs is justified.
Ol 0
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~ Page 4 Mr. Ed Hartwig April 24, 1993 HLG-93-020, Rev.
1 We hope that. the above summary is helpful;to you. Ifyou have any questions regarding the above, please call.
H. L Gustin,,P. E.
Associate/QA Manager
/gal-P. C. Riccardella.
Associate
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