ML20141G546
| ML20141G546 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/20/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141G536 | List: |
| References | |
| NUDOCS 9705220408 | |
| Download: ML20141G546 (6) | |
Text
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REACTOR COOIANT SYSTEM BASEQ The total steam generator tube leakage limit of 300 gallons per day l
for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits
..in the event of either a steam generator tube rupture or steam line break.
~ ~ The 150 gallon per~ day leskage' limit per steam generator ensures-that steam-
-generator tube integrity.is maintained in the event of a main steam line rupture or under LOCA conditions.
l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since l
.it may be. indicative of an impending gross failure of the pressure houndary. therefore, the presence of any PRESSURE BOUNDARY LEAKAGE
-requires the 4anit to be promptly placed in COLD SHUTDOWN.
3/4. 4."1 - CHEMISTRY
.The limitations on. Reactor T.o lant system chemistry ensure that l
corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the steady state Limits
_ rovides adequate corrosion protection to ensure the structural integrity p
, of the Reactor coolant system over the life of the plant. The associated effects of exceeding the oxygen,4:hloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the steady state Limits, up to the Transient Limits, for the specified limited time l.
intervals without having a significant effect on the structural integrity l 4 of the Reactor Coolant system. The ting interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the' contaminant concentrations to within the steady 3 tate Limits.
The surveillance requirements provide adequate assurance that con-
-centrations in excess of the limits will be detected in sufficient time to take corrective action.
o 3/4.4.8 SPECIf'IC ACTIVITY r
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_.The limitations on the. specific activity of the. primary coolant
-- ensure that the resulting 2-hour doses at the site boundary will not l
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- -exceed an appropriately small fraction of Part 100 limits following a
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ARRANSAS - UNIT 2
'S 3/4 4-4 Amendment No.'184 1
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.184TO l
FACILITY OPERATING LICENSE NO. NpF-6 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NO. 2 DOCKET NO. 50-368 i
1
1.0 INTRODUCTION
By letter dated November 26, 1996, and as suppl 6mented by letter dated February 12, 1997, Entergy Operations, Inc. (the licensee), submitted a l
request to change the Technical Specifications (TS) for Arkansas Nuclear One, l
l Unit 2 (ANO-2).
The proposed changes incorporate references to a new l
Combustion Engineering (CE) topical report describing steam generator (SG)
]
tube sleeves, delete references to the previous CE topical report, change the l
plugging limit for a CE sleeve to 29% of the nominal sleeve wall thickness, require post weld heat treatment (PWHT) of sleeve welds, and reduce the allowable primary to secondary leakage through any one steam generator to 150 gallons per day (gpd). The information provided in the letter dated February 12, 1997, provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
I By letter dated January 26, 1993, the NRC staff issued license amendment number 142 for ANO-2 allowing the repair of steam generator tubes using CE i
designed welded sleeves as described in the CE topical report CEN-601-P, Revision 01-P, "ANO-2 Steam Generator Tube Repair Using Leak Tight Sleeves,"
dated July 1992. The requested TS changes will reference a new generic topical report, CEN-630-P, Revision 01, " Repair of 3/4" 0.D. Steam Generator il Tubes Using Leak Tight Sleeves," dated November 1996. The latter report is not significantly different from CEN-601-P other than it addresses issues j
identified previously at Prairie Island Unit 1 (PI-1) associated with indications detected in weld joints of CE sleeves resulting from inadequate l
cleaning.
The report also revises the sleeve plugging limit and renders optional the PWHT of the sleeve welds.
Because the bulk of the technical and regulatory issues for the present request are identical to those reviewed in the previous Safety Evaluation (SE) described in our January 26, 1993 letter, l
this SE discusses only those issues warranting revision, amplification or inclusion based on current experience.
Details of prior staff evaluations of CE sleeves may also be found in the SEs for Waterford Steam Electric Station, Unit 3, Docket No. 50-382, dated December 14, 1995; Byron Nuclear Power Station, Units 1 and 2 and Braidwood 9705220408 970520
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PDR ADOCK 05000368 P
. Nuclear Power Station, Units 1 and 2, Docket Nos. 50-454, 50-455, 50-456, and 50-457, dated April 12, 1996; and Zion Nuclear Power Station, Units 1 and 2, Docket Nos. 50-295 and 50-304, dated October 29, 1996.
These evaluations apply to the proposed ANO-2 license amendment.
2.0 BACKGROUND
The two proposed CE sleeve types are an expansion transition zone (ETZ) sleeve and a tube support (TS) sleeve. An ETZ sleeve is designed to restore the
' portion of a tube in the vicinity of the top of the SG tubesheet. A TS sleeve can be used to span a support plate elevation or be used on a freespan section of tube. The sleeve material is a nickel-iron-chromium alloy, alloy 690, a l
Code approved material (ASME SB-163), incorporated in ASME Code Case N-20.
The CE sleeves are installed using gas tungsten arc welding to join the sleeve to the parent tube at the upper (free span) end of the ETZ sleeve and at both ends of a TS sleeve.
The lower ETZ sleeve tube joint is hard-rolled into the tubesheet below the expansion zone.
The centerline of the welds form the pressure boundary transition between the sleeve and the tube.
The weld joint is the subject of the modifications to the installation processes described in the new topical report.
During the Spring 1996, refueling outage at PI-1, roughly 60 upper weld joints in CE sleeved tubes had eddy current testing (ET) indications. Discovery of most of the indications was the result of the licensee employing a new, more sensitive ET probe for its periodic inspection of SG tubes.
Tube / sleeve assemblies were removed from the SGs for metallurgical examination and root cause determination.
It was found that the ET indications were due to entrapped oxides and/or weld suckback within the sleeve to tube weld.
The cause of these weld defects was traced to a previously revised tube cleaning procedure.
As a result of the metallurgical examination, the tube cleaning procedure was revised and revised post cleaning visual inspections (VT) were adopted.
The initial weld acceptance inspection, an ultrasonic test (UT), was revised to give greater sensitivity. As an added measure, the initial baseline ET, normally used only as reference for later periodic reinspection, was modified to supplement the UT as part of the initial weld acceptance inspection.
All of these refinements to the sleeving procedure were confirmed using a large number of laboratory samples and field mock-ups.
These modifications were incorporated into a new generic topical report, CEN-630-P, referenced above and are discussed in more detsil in the following section.
3.0 DISCUSSION Experience with all types of S3 tube sleeves has led to several areas of concern outside the scope of basic sleeve design and qualification discussed I
in previous SEs.
These include weld preparation, weld acceptance inspections, sleeve plugging limits,' service life predictions for sleeved SG tubes, and primary-to-secondary leakage limits.
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3.1 Weld Preparation Prior to performing any weld, the surface of the metal (s) to be welded must be cleaned. For sleeve installation, the inner diameter of the parent tube at the desired weld location must be cleaned of service induced oxides.
For the CE sleeving process, this is accomplished using motorized wire brushes.
Based upon the metallurgical findings, CE revised the cleaning method to ensure optimum removal of service induced oxides. The revised cleaning procedure entailed some equipment changes. More significantly, from a quality assurance standpoint, a 100 percent VT of the cleaning process was instituted.
After the wire brush cleaning step, every tube is given a VT using a remote fiber optic camera system to confirm that adequate surface cleaning has been l
accomplished.
CE advises the 100 percent VT is an interim step until enough field experience is gained to consider adoption of a statistical sampling plan in the future.
'3.2 Weld Acceptance Inspections For compliance with the Code and regulatory requirements, initial and periodic examinations of steam generator tubes and sleeves are performed.
Sleeve welds were historically accepted based on VT and UT examinations.
ET was used for an initial baseline inspection for comparison with later required periodic inspections.
The reason for the different types of nondestructive examinations (NDE) being used for initial acceptance versus periodic reinspection is due to the differences between potential flaws from initial installation defects and service induced degradation. The different NDE techniques have normally been better suited.for the respective types of anticipated flaws.
The PI-l event suggested that the current initial acceptance examinations (VT and UT) may not be sufficient in every circumstance.
As a result, the weld acceptance NDE was modified to include:
100 percent UT with an enhanced digitized amplitude system I
100 percent ET using the Plus Point probe The PI-l event indicated that cleaning the parent tube prior to welding is a critical step in forming a defect-free sleeve to tube weld. Thus the new CE topical report requires a 100 percent VT of the parent tube after cleaning.
The original UT procedure was based upon the absence of a mid-wall reflection.
l In that procedure, the sleeve outside diameter w.all reflection was readily apparent beyond the fusion zone of the weld, thus signifying lack of fusion with the parent tube. When fusion existed, the mid-wall reflection (mid-wall of the fused sleeve and tube combination) would not appear since no interface would exist. The PI-l event led CE to discover that lack of fusion caused by axially oriented oxide inclusion from a poorly cleaned weld would not be detected since the oxides did not cause a large sound reflection.
In the enhanced UT procedure, the back' wall signal from the outside of the parent tube is also monitored for presence in the fused area.
Additionally,
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the back wall signal strength is examined for excessive attenuation.
Attenuation beyond the normal amount can be interpreted, along with other signal artifacts. as either a weld that is too narrow or one with inclusions or patches of uniused material. The mcdified UT procedure was extensively tested on laboratory produced welds containing a variety of inclusion / lack of fusion defects.
Samples were destructively examined and the metallurgical sections compared with the UT results. Comparison of results demonstrated the revised UT procedure was highly reliable, and that no significant defects could remain undetected by the enhanced UT procedure.
ET with the plus point probe is now part of the sleeve weld acceptance criteria. The PI-I event led CE to discover that weld suckback and circumferential1y oriented oxide inclusions from a poorly cleaned weld would not be detected by UT.
CE has shown the plus point probe reliably detects the various process-induced weld defects including blowholes, weld suckback and circumferential1y oriented oxide inclusions.
CE has also shown the ET can reliably locate the position of the defect with respect to the weld centerline which is considered the pressure boundary.
ET indications located above the weld centerline that meet UT requirements can be left in service.
Any ET indication found below the weld centerline requires the tube to be plugged.
For future sleeve inspections, the licensee will follow EPRI guidelines for determi6ation of inspection scope and expansion criteria. The licensee will use Electric Power Research Institute (EPRI) "PWR Steam Generator Tube Examination Guidelines" Appendix G qualified personnel and Appendix H qualified ET techniques.
3.3 Sieeve Plugging Limits The sleeve minimum acceptable wall thickness is determined using the criteria i
of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steem Generator i
Tubes" and ASME Code Section III allowFe stress values and pressure stress equations. According to RG 1.121 criteria, an allowance for NDE uncertainty j
and postulated operational growth of tube wall degradation within the sleeve must be accounted for when using NDE to determine sleeve plugging limits.
Therefore, a conservative tube wall combined allowance for postulated degradatior growth and eddy current uncertainty of 20% throughwall per cycle was assumed for the purpose of determining the sleeve plugging limit. The sleeve plugging limit, which was calculated based on the most limiting of normal, upset, or faulted conditions for 3/4-inch outside diameter steam generator tubes in CE designed generators, was determined to be 49% of the sleeve nominal wall thickness based on ASME Code minimum material properties l
in accordance with staff positions.
Removal of tubes and/or sleeves from service when degradation reaches a plugging limit of 29% provides assurance the minimum acceptable wall thickness will not be violated during the next subsequent cycle of operation.
t
. 3.4 Post Weld Heat Treatment Accelerated corrosion tests confirm a PWHT significantly improves the intergranular stress corrosion cracking resistance of the alloy 600 parent tube material in the weld zone. The licensee committed to performing a PWHT of the welded joints in accordance with the CE topical report and NRR staff position. This commitment is reflected in the TS.
3.5 Primary to Secondary Leakage Limits With respect to the staff position regarding primary to secondary leakage limits, the licensee proposes to change its TS adopting a 150 gpd per SG
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leakage limit.
3.6 Technical Specification Changes The staff finds acceptable the following proposed changos to the plant TS 4.4.5.4 and 3.4.6.2 and associated bases.
1.
The defi:.ition of the sleeve plugging limit is modified to incorporate a revised plugging limit of 29%.
2.
The CE sleeve installation reference document is changed to indicate CE sleeves will be installed as described in CE report CEN-630-P.
3.
A PWHT of the sleeve welds as described in CE report CEN-630-P shall be performed.
4.
The allowable primary to secondary leakage through any one SG is reduced to 150 gallons per day.
4.0 TECHNICAL CONCLUSIONS The staff concludes the proposed dMving repairs, as described in the new CE sleeve topical report, can be acct,br.Ished to produce sleeved tubes of acceptable metallurgical properties, structural integrity, leak tightness and corrosion resistance.
The staff also finds acceptable the proposed preservice and future inspection methods for examining the tube / sleeve assemblies.
The NRC staff concludes the repair of SC tubes using welded sleeves designed by CE is acceptable, as implemented through appropriate TS changes that
- 1) modify the sleeve plugging limit, 2) reference the updated CE topical report, 3) perform a PWHT of '
welded joints, and 4) modify the TS requirements to incorporate a primary to secondary leakage limit of 150 gpd maximum per SG.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was n'otified of the proposed issuance of the amendment. The State official had no comments.
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6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes' a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR -
- Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the. amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in in6ividual or cumulative occupational radiation exposure. The Commission has previously issued a pro-3 posed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (61 FR 64376). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR -51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
7.0 CONCLUSION
The. Commission has concluded, based on the considerations discussed above, 5
that:
(1) there is reasnnable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal _ Contributors:
S. M. Coffin G. P. Hornseth Date:
May 20. 1997 i
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