ML17313B051
| ML17313B051 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/05/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17313B050 | List: |
| References | |
| NUDOCS 9908120128 | |
| Download: ML17313B051 (22) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON) 0 C 2055&0001 SAFETY EVALUATIONBY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENTNO. 120 TO FACILITYOPERATING LICENSE NO. NPF%1, AMENDMENTNO. 120 TO FACILITYOPERATING LICENSE NO. NPF-51 ANDAMENDMENTNO. 120 TO FACILITYOPERATING LICENSE NO. NPF-74 ARIZONAPUBLIC SERVICE COMPANY ET AL.
'PALO VERDE NUCLEAR GENERATING STATION UNITS 1 2 AND 3 DOCKET NOS. STN 50-528 STN 50-529 AND STN 50-530
1.0 INTRODUCTION
By application dated May 23, 1997, as supplemented by letters dated September 27, 1998, and May 26, 1999, the Arizona Public Service Company (APS or the licensee) requested changes to the Technical Specifications (TS) for the Palo Verde Nuclear Generating Station (Palo Verde),
Units 1, 2, and 3. APS submitted this request on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edison Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority. The proposed changes would allow the installation of ABB Combustion Engineering (ABB-CE) leak-tight sleeves in defective steam generator tubes as a tube repair method.
2.0 BACKGROUND
The ABB-CE sleeves consist of a tubesheet sleeve design and a tube support sleeve design.
A tubesheet sleeve is designed to repair the degraded portion of a tube in the vicinity of the top of the tubesheet.
A tube support sleeve is designed to repair the degraded freespan or support plate region of a tube. The licensee-proposed sleeve repair method is based on the report, "Repair of 3/4-inch O. D. [Outer Diameter] Steam Generator Tubes Using Leak Tight Sleeves,"
CEN-630-P, Revision 02, June 1997 (Proprietary information. Not publicly available. A nonproprietary version of CEN-630 was submitted to NRC on September 16, 1997).
The staff has approved thh use of similarly designed sleeves in U.S. nuclear plants.
The staff review of the licensee's submittal is therefore focused on those issues warranting revision, amplification, or inclusion based on recent field experience.
Details of prior staff evaluations of ABB-CE leak-tight sleeves may be found in the safety evaluations for Waterford Steam Electric Station, Unit 3, docket number 50-382, dated December 14, 1995; Byron Nuclear Power Station, Units 1 and 2 and Braidwood Nuclear Power Station, Units 1 and 2, docket numbers 50-454, 50%55, 50-456, and 50-457, dated April 12, 1996; Kewaunee Nuclear Power Plant, docket No.
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, 50-305, dated June 7, 1997; Prairie, Island Units 1 and 2, docket numbers 50-282 and 50-306, dated November 4, 1997; Beaver Valley Unit 1
~ docket number 50-334, dated November 25, 1997; and San Onofre Units 2 and 3, docket numbers 50-361 and 50-362, dated August 26, 1998.
Previous staff evaluation of ABB-CE sleeves addressed the technical adequacy of the sleeves in the four principal areas of pressure-retaining component design:
structural requirements, material of construction, welding, and nondestructive examination.
The staff found the analyses and tests that were submitted to address these areas of component design to be acceptable.
The function of sleeves is to restore the structural and leakage integrity of the tube pressure boundary.
Consequently, structural analyses were performed for a variety of loadings iricluding design pressure, operating transients, and other parameters selected to envelop loads imposed during normal operating, upset, and accident conditions.
Stress analyses of sleeved tube assemblies were performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III. As described in detail in CEN-630-P, Revision 02, the structural integrity of the sleeve design has been investigated analytically and verified by laboratory tests of steeve mockups.
These analyses, along with the results of qualification testing.and previous plant operating experience, were cited to demonstrate that the sleeved tube assembly is capable of restoring steam generator tube integrity.
The sleeve material, thermally treated (Tl ) Alloy690, is a nickel-iron-chromium alloy. It is an ASME Code-approved material, specified in ASME SB-163,, and is incorporated in ASME Code Case N-20. The, staff has determined that the use of Alloy690 TT material is an improvement over'the Alloy600 material used in the parent tube.
Corrosion tests conducted under Electric Power Research Institute (EPRI) sponsorship confirm that Alloy690 TT material resists corrosion better than that of Alloy600. As a result of these laboratory corrosion tests, the staff has concluded that Alloy690 TT material satisfies the guidelines in Regulatory Guide (RG) 1.85, "Materials Code Case Acceptability ASME Section III, Division 1," Revision.24, dated July 1986.
The staff has approved use of Alloy690 TT tubing. in previous sleeving,applications.
For the tubesheet sleeve, the upper end of the sleeve is welded to the parent tube in the freespan region above the tubesheet and the lower end is hard-rolled into the tubesheet below the expansion zone.
For the tube support sleeve, both ends of the sleeve are welded to the parent tube. The welding process uses automatic autogenous gas.tungsten arc welding which was qualified and demonstrated during laboratory tests by full-scale mock-ups.
Qualification of the welding procedures and welding equipment operator was performed in accordance with the specifications of the ASME Code,Section IX.
The staff considers sleeves to be a long-term repair but not a repair with unlimited service life.
The welding of the steeve to the tube may-create new locations-susceptible to stress corrosion cracking and the time for the initiation of service-'induced. degradation in sleeve-tube assemblies is difficultto quantify. The staff finds the accelerated corrosion tests of sleeve-tube assemblies, conducted by the vendors to'redict service life, not reliable for deterministic predictions.
In order to get a more realistic evaluation of the degradation, the licensees inspect'a sample of sleeves at each outage to ensure that any sleeve degradation is detected and addressed early.
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, This results in a better and more accurate prediction of the sleeve life. The inservice inspection requirements for the sleeve inspection are discussed further in Section 3.2 of this safety evaluation. The staff considers the corrosion tests coupled with the inservice inspection of the sleeves at each outage to give a viable indicator of potential performance and therefore acceptable.
3.0 EVALUATION Experience with all types of steam generator tube sleeves'has revealed certain issues that need to be evaluated in addition to sleeve design and qualification as discussed in previous NRC safety evaluations.
These issues, involve weld preparation, weld acceptance inspections, inservice inspection expansion criteria, sleeve plugging limits, post-weld heat treatment, and primary-to-secondary leakage limits, which are discussed below.
3.1 Weld Pre aration andAcce tance Ins ections During its spring 1996 refueling outage, the licensee of a domestic nuclear plant detected eddy current test (ET) indications in. about 60 weld joints in ABB-CE leak-tight sleeves.
This finding was the result of using a new, more sensitive ET probe. The ET indications were caused by entrapped oxides and/or weld shrinkage within the sleeve-to-tube weld. The cause of these weld defects was traced to an inadequate tube cleaning process.
Although the defects did not significantly impair the. structural integrity (strength) of the welds and did not cause leakage,'they did increase the probability of leakage.
In a separate case, during an installation of welded sleeves in another domestic nuclear plant, weld zone indications were identified visually but were not detected by either ET or ultrasonic testing (UT). These findings pointed to the inadequacy of previous sleeve installation and inspection.
ABB-CE has revised its weld preparation procedures and incorporated these changes in CEN-630-P, Revision 02. Prior to installing a sleeve, the inner surface of the, parent tube at the desired weld location is cleaned of service-induced oxides using motorized wire brushes.
ABB-CE specifies that after surface cleaning, every repaired tube be visually inspected to confirm adequate surface cleaning. ABB-CE advises that the visual'inspection of every tube is an interim measure until sufficient field experience is gained to consider adoption of statistical sampling in the future.
ABB-CE has also revised its procedures for weld acceptance. inspection. The initial weld acceptance inspection, performed by UT, was revised to give greater sensitivity. An optional visual inspection, the VT-1 inspection process specified in ASME Code Section XI, was added'o the inspection procedure. The initial baseline ET, normally used only as a reference for future inservice inspections, was modified to supplement the UT as a part of the weld acceptance inspection.
Allof these refinements to the sleeve inspection were confirmed using a large number of laboratory samples and field mockups.
The refinements have been incorporated into CEN-630-P, Revision 02, and are discussed in detail below.
The original UT procedure for the sleeve weld joint was based upon the absence of a mid-wall reflection.
In an. acceptable sleeve-to-tube weld, the mid-wall reflection (mid-wall of the fused sleeve and tube) would not appear because no interface would exist. Previous field experience
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The lack of fusion was caused by axially oriented oxide inclusions from an inadequately prepared tube surface.
In the improved UT procedure, the back wall signal from the outside of the parent tube is also monitored for presence in the fused area.
Additionally, the back wall signal strength is examined for excessive attenuation.
Attenuation beyond the normal signal strength, along with other signal artifacts, can be used to detect unacceptable welds. ABB-CE tested the enhanced UT procedure and demonstrated that the revised UT procedure is reliable. As stated in CEN-630-P, Revision 02, use of the Plus Point probe is now part of the sleeve weld acceptance criteria. ABB-CE has shown that the Plus Point probe reliably detects these process-induced weld defects and blowholes.
The licensee stated that it willperform the required visual inspection after tube cleaning in accordance with CEN-630-P, Revision 02. It willalso conduct UT and ET examinations after the completion of the sleeve-to-tube weld for all installed sleeves in accordance with CEN-630-P, Revision 02.
In addition, the licensee willperform a VT-1 inspection of each sleeve-to-tube weld until sufficient data has been obtained with UT and ET techniques to show that these techniques are capable of detecting and resolving uncertainties in the weld joint.
Accordingly, the staff finds the proposed weld inspection method acceptable.
3.2 Inservice Ins ection Re uirements For inservice inspection of sleeved tubes, the licensee has proposed to perform an initial inspection of 20% of sleeves at each refueling outage.
This initial sample size is more than the initial tube sample size of 3% required by the current TS. The minimum sample requirements for tube inspections are established to assess the overall condition of steam generator tubing.
The licensee's proposed inspection sampling for sleeved tubes is consistent with the current industry guidance for steam generator sleeve examinations as specified in EPRI report, "Steam Generator Examination Guidelines," TR-107569, Revision 5 (Proprietary. A nonproprietary document is available). The licensee's proposed inspection sampling for sleeved tubes is also consistent with sleeve inspection sampling plans previously approved by the staff and detailed in Section 2.0, paragraph 2, of the safety evaluation.
Additionally, the proposed TS will require additional tubes be inspected, if warranted by the tube inspection results.
In view of the above, the staff considers the inservice inspection program for the sleeved tubes adequate to detect degradation in them and, therefore, acceptable.
3.3 Sleeve Plu in Limit The sleeve plugging limitis defined in the proposed TS as the imperfection depth in the sleeve at or beyond. which the sleeved tube shall be removed from service. The sleeve plugging limit is calculated from the minimum acceptable sleeve wall thickness to maintain structural integrity..
RG 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," and ASME Code Section III provide guidance on the calculations.
In addition to structural consideration, RG 1.121 also suggests that an allowance for nondestructive evaluation (NDE) uncertainty and postulated growth of degradation be accounted for in the sleeve plugging limitwhen using NDE to evaluate sleeve degradation.
The licensee assumes a 10% allowance for ET uncertainty and a 10% allowance for degradation growth per cycle in its calculations.
The licensee calculated a minimum acceptable wall thickness of 54.67% due to structural consideration.
After deducting a total allowance of 20%, the licensee specified a sleeve plugging limitof 35% of sleeve wall
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'hickness.
Thelicenseehasproposedthisrequirementin TS5.5.9.4.a.6.
Thestafffindsthat the sleeve plugging limitsatisfies the recommendations in RG 1.121, and therefore, is acceptable.
3.4 Post-Weld Heat Treatment Residual stress is a contributor to stress corrosion cracking in steam generator tubing. The welding of the sleeve to the parent tube willintroduce residual stresses in both the sleeve and the tube. These stresses may increase the susceptibility of the welded joints to stress corrosion cracking. A post-weld heat treatment (PWHT) can reduce these stresses and thus may reduce the likelihood of cracking within a welded joint. ABB-CE recommends that a PWHT be a part of the sleeve installation process.
The licensee willfollowthe recommendation in CEN-630-P, Revision 02, in regard to PWHT of the welded joints.
The staff considers the actions planned by the licensee in regard to PWHT adequate to reduce the residual stresses which may in turn reduce the likelihood of cracking within a welded joint and, therefore, acceptable.
3.5 Prima
-to-Seconda Leaka e Limit Leak resistance of the sleeve has been demonstrated through laboratory tests.
Bounding calculations and laboratory tests have verified that, should leakage develop in the welded or rolled joints of sleeved tubes, it would not exceed 1 gallon per minute (gpm) and, thus, the 10 CFR Part 100 requirements, for radiological release would not be affected, even under the most severe postulated conditions.
In addition, the licensee has proposed to modify the current primary-to-secondary leakage limitof 500 gallons per day through any one steam generator to the more stringent 150 gallons per day through any one steam generator.
This modification is stated in TS 3.4;14.d and is consistent with the operational leakage limitaccepted by the staff in other alternate repair reviews. The staff has determined that the primary-to secondary leakage limits verified by laboratory tests and the licensee's proposal to modify the current primary-to secondary leakage limits to more restrictive and stringent values are adequate to keep the radiological releases to within the 10 CFR Part 100 requirements and are therefore acceptable.
3.6 Pro osedTS Chan es In order to implement sleeving of the degraded tubes in Palo Verde steam generators, the licensee has proposed the following changes to the plant TS. The changes (other than editorial changes) are summarized below:
TS 3.4.14.d The primary-to-secondary leakage is limited to 150 gallons per day through any one steam generator.
TS 5.5.9 The Steam Generator Tube Surveillance Program is revised to include sleeves.
TS 5.5.9.1 The Steam Generator Sample Selection and Inspection section is revised to include the inspection of sleeves.
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~ TS 5.5.9.2 TS 5.5;9.2.a and 5.5.9.2.b are revised to include sleeve inspection criteria in addition to the existing tube inspection criteria. Table 5.5.9-3 specifies sleeve sample inspection and associated expansion criteria and is referenced in this section.
TS 5.5.9.3 The Inspection Frequency section is revised to include the inspection frequency for sleeves.
A requirement is also added to perform a preservice inspection of tubes that have been repaired by sleeving.
TS 5.5.9.4 The Acceptance Criteria section is revised to include sleeves in the definitions of Defect, Degradation, Degraded Sleeve, Plugging or Repair Limit, Preservice Inspection, Sleeve Inspection, and Tube. Repair or Sleeve.
The plugging limitis specified to be 35% of sleeve wall thickness.
TS 5.5.9.4.11 CEN-630-P, Revision 02, is referenced in this section.
TS 5.6.8 The Steam Generator Tube Inspection Report section requires that specific reports be submitted to NRC when a tube is repaired by sleeving.
The changes contained in the TSs are consistent with the preceding evaluation of the sleeving amendment.
The staff concludes that the licensee's proposal on setting the limiton primary-to-secondary leakage, revision of the Steam Generator Tube Surveillance Program to include sleeves, inclusion of the sleeves in the Steam Generator Sample Selection and Inspection section, including the sleeve inspection criteria in addition to the existing tube inspection criteria, and revision of the existing inspection frequency for the sleeves are acceptable.
The staff also reviewed the accompanying editorial changes proposed by the licensee and concludes that they have no substantive effect on plant operation (repagination, changes to section numbers) and are therefore acceptable.
Based on the foregoing, the staff concludes that the licensee may incorporate the proposed changes into Palo Verde Units 1, 2, and 3 TSs.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arizona State officialwas notified of the proposed issuance of the amendments.
The State official had no comments.
5.0 ENVIRONMENTALCONSIDERATION The amendments change requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The NRC staff has determined that the amendments involve 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 individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (64 FR 32285). Accordingly, the amendments meet the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). The amendments
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.Accordingly, with respect to such changes, the amendments meet the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public willnot be endangered by operation in the proposed manner, (2) such activities willbe conducted in compliance with the Commission's regulations, and (3) the issuance. of the amendments willnot be. inimical to the common defense and security. or to the, health and safety, of-the public.
Principal Contributor: J. Tsao Date: August 5, 1999
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June 25, 1999 Mr. James M. Levine S'enior Vice President, Nuclear Arizona Public Service Company Post Office Box 53999 Phoenix, AZ 85072-3999
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANTIMPACT RELATEDTO ISSUANCE OF EXEMPTION TO 10 CFR 50.71(e)(4) (TAC NOS. MA2159, MA2160, AND MA2161)
Dear Mr. Levine:
Enclosed is a copy of the Environmental Assessment and Finding of No Significant Impact related to your application for exemption dated June 9, 1998, as supplemented by letter dated December 21, 1998.
In these letters, you requested an exemption for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3, from a requirement of 10 CFR 50.71(e)(4) to submit updates to the Updated Final Safety Analysis Report (UFSAR) annually or 6 months after each refueling outage.
In addition, you requested that this exemption also apply to (1) revisions made to the quality assurance program (which has been incorporated into the UFSAR), (2) the safety evaluation summary reports for facilitychanges made under 10 CFR 50.59, and (3) the reports of changes to the Technical Specification Bases.
This assessment is being forwarded to the Office of the Federal Register for publication.
Sincerely, ORIG.
SIGNED BY Mel:B. Fields, Project Manager, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of.Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530
Enclosure:
Environmental Assessment cc w/encl: See next page DISTRIBUTION "Docket File OGC PUBLIC ACRS PDIV-2 Reading SRichards (clo)
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Mr. James M. Levine Senior Vice President, Nuclear Arizona Public Service Company Post Office-Box 53999 Phoenix, AZ 85072-3999
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - ENVIRONM AL ASSESSMENT AND FINDING OF NO SIGNIFICANTIMPACT LATEDTO ISSUANCE OF EXEMPTION TO 10 CFR 50.71(e)(4) (TAC S. MA2159, MA2160, AND MA2161)
Dear Mr. Levine:
Enclosed is a copy of the Environmental Assessment and Fi ing of No Significant Impact related to your application for exemption dated June 9, 19, as supplemented by letter dated December 21, 1998.
In these letters, you requested an emption for the Palo Verde Nuclear Generating Station, Units 1, 2', and 3, from a require nt of 10 CFR 50.71(e)(4) to submit updates to the Updated Final Safety Analysis Repo (UFSAR) annually or 6 months after each refueling outage.
In addition, you requested that is exemption also apply to (1) revisions made to the quality assurance program (which as been incorporated into the UFSAR), (2) the safety evaluation summary reports for facilit changes made under 10 CFR 50.59, and (3) the reports of changes to the Technical Speci ation Bases.
This assessment is being forwarded t the Office of the Federal Register for publication.
Sincerely, Docket Nos. 50-52, 50-529, and 50-530
Enclosure:
Env'nmental Assessment cc w/encl:
e next page DISTRIB TION Mel B. Fields, Project Manager, Section 2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docke ile PUB C
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 25, 1999 Mr. James M. Levine Senior Vice.President, Nuclear Arizona Public Service Company Post Office Box 53999 Phoenix, AZ 85072-3999
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANTIMPACT RELATED TO ISSUANCE OF EXEMPTION TO 10 CFR 50.71(e)(4) (TAC NOS. MA2159, MA2160, AND MA2161)
Dear Mr. Levine:
Enclosed is a copy of the Environmental Assessment and Finding of No Significant Impact related to your application for exemption dated June 9, 1998, as supplemented by letter dated December 21, 1998.
In these letters, you requested an exemption for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3, from a requirement of 10 CFR 50.71(e)(4) to submit updates to the Updated Final Safety Analysis Report (UFSAR) annually or 6 months after each refueling outage.
In addition, you requested that this exemption also apply to (1) revisions made to the quality assurance program (which has been incorporated into the UFSAR), (2) the safety evaluation summary reports for facilitychanges made under, 10 CFR 50.59, and (3) the reports of changes to the Technical Specification Bases.
This assessment is being forwarded to the Office of the. Federal Register for publication.
Sincerely, Mel B. Fields, Project Manager, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530
Enclosure:
Environmental Assessment cc w/encl: See next page
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Palo Verde Generating Station, Units 1, 2, and 3 CC:
Mr. Steve Olea Arizona Corporation. Commission 1200 W. Washington Street.
Phoenix, AZ 85007 Douglas Kent Porter Senior Counsel Southern California.Edison Company Law Department, Generation Resources P.O.'ox 800
- Rosemead, CA 91770 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 40 Buckeye, AZ 85326 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Chairman, Board of Supervisors ATTN: Chairman 301 W. Jefferson, 10th Floor Phoenix, AZ 85003 Mr..David Summers Public Service Company of New Mexico 414 Silver SW, 01206 Albuquerque, NM 87102 Mr. Jarlath Curran Southern California Edison Company 5000 Pacific Coast Hwy Bldg DIN San Clemente, CA 92672 Mr. Robert Henry Salt River Project 6504 East Thomas Road Scottsdale, AZ 85251 Terry Bassham, 'Esq.
General Counsel El Paso Electric Company 123 W. Mills El Paso, TX 79901 Mr. John Schumann Los Angeles Department of Water & Power Southern California Public Power Authority P.O. Box 51111, Room 1255-C Los Angeles, CA 90051 Mr. Aubrey V. Godwin, Director Arizona. Radiation Regulatory Agency 4814 South 40 Street Phoenix, AZ 85040 Ms. Angela K. Krainik, Manager Nuclear Licensing Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034 Mr. John C.'orne Vice President, Power Generation El Paso Electric Company 2702 N. Third.Street, Suite 3040 Phoenix, AZ 85004 May 19,1999
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