ML20137D696
| ML20137D696 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/24/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137D682 | List: |
| References | |
| NUDOCS 9703260270 | |
| Download: ML20137D696 (9) | |
Text
_ _. _ _ _ _ _ _ _ _ - _ _. _ _ _ _ _ _ _. _.. _.. _. _. _. _ _. _ _ _ _. _ _ _ _ _. _ _
[presh
{
g i
UNITED STATES i
s j
NUCLEAR REGULATORY COMMISSION
/
2 WASHINGTON, D.C. 30e86 4001
\\..... p$
l t
i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
RELATED TO AMENDMENT NO. 124 TO FACILITY OPERATING LICENSE NO. NPF-2 i
i SOUTHERN NUCLEAR OPERATING COMPANY. INC.
I JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 i
DOCKET NO. 50-348 i
l
1.0 INTRODUCTION
i By letter dated December 26, 1996, as supplemented by letters dated February 6, March 7, March 21, 1997, Southern Nuclear Operating Company, Inc.,
j et al. (the licensee), submitted for staff review a license amendment to change the Technical Specifications (TS) for the Joseph M. Farley Nuclear i
Plant, Unit 1.
The licensee proposed to implement permanent voltage-based alternate repair criteria for steam generator tubes in the TS. The proposed 4
i alternate repair criteria would allow steam generator tubes having outside diameter stress corrosion cracking (00 SCC) that is predominately axially l
oriented and confined within the tube support plates to remain in service on j
the basis of bobbin coil voltage response. The NRC guidance on the alternate repair criteria is specified in Generic Letter (GL) 95-05, " Voltage-Based i
Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," August 3, 1995.
i In addition to the TS amendment, the licensee proposed the following i
additional items: (1) to include 50% of the bobbin coil indications not confirmed by rotating pancake coil in the determination of the beginning-of-l cycle voltage distribution instead of 100% as specified in GL 95-05; (2) proposed to use probability of detection that is voltage dependent, instead of a constant 60%; and (3) revising the steam line break leakage limit from 11.4 i
gpa to 20 gpm. By letter dated January 27, 1997, the staff requested from the licensee additional information related to items (1) and (2).
In the 1
licensee's February 6, 1997, response, they requested that the proposed TS l
changes be approved without approval of items (1) and (2). Also, by letter dated March 21, 1997, the licensee withdrew their request for approval of item (3).
By letters dated February 6, March 7, and March 21, 1997, the licensee submitted additional information to clarify the changes to the proposed repair l
4 criteria, which did not change the scope of the December 26, 1996, application 1
and the initial proposed no significant hazards consideration determination.
i 9703260270 970324 PDR ADOCK 05000348 i
P pop I
~. ~
~.
1 i
- l l
2.0 BACKGROUND
i Steam generator tube flaw acceptance criteria (i.e., plugging limits) are i
specified in the plant TS. The traditional strategy for achieving adequate i
structural and leakage integrity of the tubes has been to establish a minimum wall thickness requirement in accordance with NRC Regulatory Guide (RG) 1.121,
" Bases for Plugging Degraded PWR Steam Generator Tubes." Development of minimum wall thickness requirements to satisfy RG 1.121 was governed by i
l analyses assuming a uniform thinning of the tube wall. This assumed degradation mode is inherently conservative for most other forms of steam j
generator tube degradation. Conservative repair limits may lead to plugging i
tubes with adequate structural and leakage integrity for further service.
j The staff developed generic criteria for voltage-based limits for 00S0C confined within the thickness of the tube support plates. The staff published l
several conclusions regarding voltage-based repair criteria in draft NUREG-l 1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" and in a draft GL titled " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the federal Register on August 12, 1994 (59 FR 41520). On August 3, 1995, the staff issued GL 95-05 that took into consideration public comments on the i
draft GL cited above, domestic operating experience under the voltage-based i
repair criteria, and additional data made available from European nuclear power plants.
The guidance of GL 95-05 does not set depth-based limits on predominantly j
axially oriented ODSCC at tube support plate locations; rather it relies on j
empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak rate.
The staff recognizes that although the total tube integrity margins may be reduced following application of a voltage-based repair criteria, the guidance in GL 95-05 ensures structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR Part 50 and the guideline values in 10 CFR.Part 100. Since the voltage-based repair criteria do not incorporate a minimum tube wall thickness requirement, there is the possibility for tubes with through-wall cracks to remain in service.
Because of the increased likelihood of such flaws, the staff included provisions for augmented steam generator tube inspections and more restrictive operational leakage limits.
GL 95-05' specifies, in part, that: (1) the repair criteria is only applicable to predominantly axially oriented ODSCC located within the bounds of the tube support plates; (2) licensees perform an evaluation to confirm that the steam generator tubes will retain adequate structural and leakage integrity from cycle to cycle; (3) licensees adhere to specific inspection criteria to ensure consistency in methods between inspections; (4) tubes must be periodically removed from the steam generators, examined, and destructively tested to verify the morphology of the degradation and provide additional data for
t,
i structural and leakage integrity evaluations; (5) the operational leakage limit be. reduced; (6) licensees implement an operational leakage monitoring program; and (7) specific reporting requirements shall be incorporated into the plant technical specifications.
The licensee has applied for the voltage-based alternate repair criteria on an interim basis and the staff ha:; approved the licensee's interim repair criteria for the Farley Unit 1 TS as documented in license Amendment No. 95, issued on October 8, 1992; license Amendment No. 106 on April 5, 1994; and license Amendment No. 117 on September 28, 1995.
Each interim criteria amendment was approved for a specific operating cycle. The proposed permanent alternate repair criteria will replace the interim criteria and will eliminate the need for applying periodic license amendments for the tube repair i
criteria.
4 Farley Unit I uses three Westinghouse model 51 steam generators. The tubes were fabricated using mill annealed alloy 600 material.
Each steam generator has 3,388 tubes and the nominal outside diameter for each tube is 7/8 inch.
3.0 EVALUATION
~
The licensee has stated that it will comply with the guidance in GL 95-05 for its' proposed permanent alternate repair criteria.
In addition, the licensee i
has proposed to incorporate verbatim the model technical specifications in GL 95-05 into the Farley Unit 1 TS. The major issues related to the licensee's i
implementation of the alternate repair criteria are discussed below.
l 3.1 Tube Repair Limits The proposed criteria will (1) permit indications confined to within the thickness of the tube support plates with bobbin voltages less than or equal to 2.0 volts to remain in service; (2) permit indications confined to within the thickness of the tube support plates with bobbin voltages greater than 2.0 volts but less than or equal to the upper voltage limit to remain in service if a motorized rotating pancake coil probe or acceptable alternative inspection does not detect degradation; and (3) require indications confined to within the thickness of the tube support plates with bobbin voltages greater than the upper voltage limit be plugged or repaired.
The proposed lower voltage limit of 2.0 volts is based on the use of a correlation between the burst pressure and the bobbin coil voltage of pulled tube and model boiler data and is consistent with the recommended value specified in GL 95-05 for 7/8-inch steam generator tubing. The upper voltage limit is based on the lower 95 percent prediction interval of the burst pressure versus bobbin voltage correlation, adjusted for lower bound material properties evaluated at the 95 percent confidence level. This voltage is further reduced to account for uncertainty in the nondestructive examination s
. technique and flaw growth over the next operating cycle. Because licensees periodically update the burst pressure and bobbin voltage database when the destructive test data from pulled tube are available, the upper voltage limit may vary as additional data is included in the correlation.
Section 1.b.1 of Attachment 1 to GL 95-05 specifies that the repair criteria do not apply to tube-to-tube support plate intersections where the tube with degradation may potentially collapse or deform as a result of the combined j
l postulated loss-of-coolant accident and safe shutdown earthquake loadings.
Licensees should perform or reference an analysis that identifies which intersections are to be excluded. The licensee submitted an analysis, WCAP-12871', Revision 2, as a part of its application for the interim alternate repair criteria on May 28, 1993. As a result of the licensee's analysis, no tubes need to be excluded from application of the voltage-based repair criteria. The staff found the licensee's assessment acceptable.
3.2 Inspection Issues Section 3.c.3 of Attachment I to GL 95-05 specifies guidance in regard to probe wear. The licensee proposed to use an alternative to GL 95-05. The industry approach, developed through the Nuclear Energy Institute, is such that if any of the probe wear standard signal amplitudes prior to probe i
replacement exceed the i 15 percent limit, all tubes having indications with i
voltage responses measured at 75 percent or greater of the lower repair limit must be reinspected with a bobbin probe satisfying the i 15 percent wear standard criterion.
The voltages from the reinspection should be used as the basis for tube repair. The NRC staff completed a review of the Nuclear Energy Institute proposed alternative method and concluded that the approach is acceptable as discussed in the letter from Brian Sheron of the NRC to Alex Marion of the Nuclear Energy Institute dated March 18, 1996. Therefore, the licensee's proposal to follow the industry approach to address bobbin coil probe wear is acceptable.
In the laboratory and field studies supporting the alternative probe wear criteria, the correlation of worn probe voltages with new probe voltages shows that for all significant voltage levels, the worn probe voltages are never less than 75% of the new probe voltage as discussed in the letter from Alex Marion of the Nuclear Energy Institute to Brian Sheron of the NRC dated January 23, 1996. However, in a 90-day inspection report from Byron Unit 1 dated September 9, 1996, a~ comparison made between the worn probe voltage and the new probe voltage resulted in a.few indications where the worn probe voltage was substantially less than 75% of the new probe voltage. The licensee for Byron Unit I evaluated these indications and concluded that the crit ~ ria to retest tubes with worn probe voltages above 75% of the repair e
limit is adequate and generally conservative due to the average trend for worn probe volts to exceed new probe voltages. Comparison of the actual and projected end-of-cycle voltages did not show anything unusual attributable to the alternate probe wear criteria. The staff _ concludes that the y
r.--
,m- - - -.
j-i j i l
l aforementioned probe wear results do not indicate an immediate need to modify the industry alternative probe wear criteria. However, the staff will continue to monitor the 90-day inspection reports of licensees using this
{
approach to probe wear, d
Section 3.b of Attachment 1 to GL 95-05 specifies guidance for tube inspection i
using the rotating pancake coil. The licensee stated that.it may use a i
motorized rotating coil probe, e.g., the + Point coil, for dispositioning bobbin coil indications. The staff encourages licensees to use the most sensitive inspection techniques available and as such this proposal is l
acceptable, i
j 3.3 Structural and Leakage Integrity Assessments The staff guidance for the implementation of the voltage-based repair criteria i
focuses on maintaining tube structural integrity during the full range of i
normal, transient and postulated accident conditions with adequate allowance i
for eddy current test uncertainty and flaw growth projected to occur during i
the next operating cycle. Tube structural limits based on RG 1.121 criteria l
require maintaining a margin of safety of 1,43 against tube failure under postulated accident conditions and maintaining a margin of safety of 3 against i
burst during normal operation. Because GL 95-05 addresses tubes affected with ODSCC confined to within the thickness of the tube support plate during normal operation, the staff concluded that the structural constraint provided by the l
tube support plate ensures all tubes to which the voltage-based criteria j
applies will retain a margin of 3 with respect to burst under normal operating i
conditions. For a postulated main steam line break accident, however, the l
tube support plate may displace axially during steam generator blowdown such t
that the ODSCC affected portion of the tubing may no longer be fully constrained by the tube support plate. Accordingly, it is appropriate to consider the 00 SCC affected regions of the tubes as free standing tubes for the purpose of assessing burst integrity under postulated main steam line l
break conditions.
In order to confirm the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the i
conditional burst probability and the total primary-to-secondary leak rate from an affected steam generator during a postulated main steam line break
}
event. To complete GL 95-05 prescribed assessments, the licensee proposes to j
follow the methodology described in WCAP-14277, Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for 00 SCC at TSP Intersections," dated December 1996.
Based on the staff's detailed review of WCAP-14277, Revision 1, including performance of confirmatory calculations, the staff finds the methodology acceptable.
GL 95-05 specifies that the structural and leakage integrity assessments should use the latest data from destructive examinations of tubes removed from licensees' steam generators. The licensee stated that the latest NRC-approved i
i database, using the NRC-approved data exclusion criteria, will be applied to i
the tube integrity evaluations.
For the upcoming Farley Unit 1 inspection and i
i a
b 4
t 2,
i GL 95-05 specified calculations, the licensee will use the database forwarded to the NRC by Duquesne Light Company for Beaver Valley Unit 1, dated March 27, 1996. The database contains the most currently available tube pull i
data from industry and also satisfies the exclusion criteria specified in GL 95-05. Therefore, the staff finds that the database submitted by Duquesne i
Light Company is acceptable for the GL 95-05 calculations for the upcoming l
Farley Unit 1 inspection.
i For the long-ters, Nuclear Energy Institute has developed a protocol for updating the steam generator degradation database. The staff will review the i
adequacy of the protocol.
Pending the implementation of an NRC-approved 1
process for updating a generic industry database for steam generator tube degradation, the licensee will provide the NRC with the database it intends to use prior to each refueling outage. The database will include the data from tubes that have been pulled and tested up to 2 months before the plant outage.
The staff finds the licensee's proposal acceptable.
GL 95-05 specifies an alternative for licensees to calculate the primary-to-secondary leakage and probability of tube burst given a main steam line break
)
using the projected end-of-cycle voltage distribution.
The licensee will perform the calculations on the basis of the projected end-of-cycle distributions.
In the event that the growth rate determinations cannot be completed before returning the steam generators to service, the licensee will use the actual end-of-cycle distributions as allowed in Section 2.c of Attachment I to GL 95-05. The licensee stated that even if the calculation made before returning the steam generators to service is based on the actual measured voltage distribution, the calculation based on the projected end-of-cycle voltage distribution will be submitted to the NRC in the 90-day report following the outage.
This approach is consistent with Section 6.b.(c) of Attachment I to GL 95-05 and is acceptable.
3.3.1 Conditional Probability of Burst The licensee will use the methodology described in Revision 1 of WCAP-14277 for performing a probabilistic analysis to quantify the potential for steam generator tube ruptures given an main steam line break event. The results of 4
the probabilistic analysis will be compared to a threshold value of 1x10'8 per cycle in accordance with GL 95-05. This threshold value provides assurance that the probability of burst is acceptable considering the assumptions of the caltdation and the results of the staff's generic risk assessment for steam generators contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value indicates ODSCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation assumed and evaluated as acceptable in NUREG-0844. The NRC staff concludes the licensee's proposed methodology for calculating the conditional burst probability is consistent with the guidance in GL 95-05 and is acceptable.
m a
.,+,
w.
g
l I
4 i l l
i 3.3.2 Accident Leakage The licensee will use the methodology described in Revision 1 of WCAP-14277 for calculating the steam generator tube leakage from the faulted steam generator during a postulated main steam line break event. The model consists of two major components:
(1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage model); and (2) a model predicting leak rate'as a function of voltage, i
given that leakage occurs (i.e., the conditional leak rate model). The staff concludes that the licensee's proposed methodology for calculating the tube leakage is consistent with the guidance in GL 95-05 and is acceptable.
3.3.3 Primary-to-Secondary Leakage During Normal Operation i
When the voltage-based repair criteria is implemented, tubes may have or may i
develop through-wall or near through-wall cracks during an operational cycle, thus creating the potential for primary-to-secondary leakage during normal operation, transients, or postulated accidents.
Postulated accident leak rates were discussed previously.
The staff concludes adequate leakage integrity during normal operation is reasonably assured by the TS limits on allowable primary-to-secondary leakage.
GL g5-05 specifies the operational leakage limits of the plant TS should be
]
reduced to 150 gallons per day. Farley Unit i TS currently limit the primary-to-secondary leakage through one steam generator to 150 gallons per day. This requirement is consistent with the guidance in GL 95-05 and is, therefore, acceptable.
l 3.4 Degradation Monitoring To confim the nature of the degradation occurring at the tube support plate elevations, tubes are periodically removed from the steam ger.erators for destructive tests. The test data from removed tubes can confirm that the nature of the degradation observed at these locations is predominantly axially oriented 00 SCC, provide data for assessing the reliability of the inspection methods, and supplement the existing databases (e.g., burst pressure, probability of leakage, and leak rate). GL 95-05 specifies that at least two tube be removed from steam generators with the objective of retrieving as many intersections as practical (minimum of four intersections) during the plant steam generator inspection outage preceding initial application of the voltage-based repair criteria. On an ongoing bases, additional tube specimen removals (minimum of two intersections) should be obtained at the first refueling outage following 34 effective full power months of operation or at the maximum interval of three refueling outages after the previous tube pull.
Alternatively, the licensee may participate in an industry-sponsored tube pull program endorsed by the staff as described in GL 95-05.
~
i i
j The licensee has removed at least four tubes, including the required number of intersections, from the Unit I steam generators for burst and leak rate j
testing and metallographic examination as a part of the interim repair criteria. The metallurgical examination confirmed that the degradation mechanism for the indications at the tube support plates was predominantly axially oriented ODSCC. For the permanent alternate repair criteria, the licensee stated that it will comply with the tube pull guidance in GL 95-05.
The staff concludes that the licensee satisfies the tube removal guidance of j
3.5 Technical Specification Changes j
The proposed amendment revised TS 3/4.4.6, " Steam Generators" and associated Bases section as part of implementing the voltage-based repair criteria for steam generator tubes. Specifically, the licensee changed the following Surveillance Requirements sections in the TS: TS 4.4.6.2, " Steam Generator Tube Sample Selection and Inspecton;" TS 4.4.6.4, " Acceptance Criteria;" and TS 4.4.6.5, " Reports." The changes incorporate the methodology of calculating the upper voltage repair limit and mid-cycle repair limits. -The licensee also changed TS Bases Section 3/4.4.6, " Steam Generators," consistent with these changes as stated above. The staff concludes that the proposed TS changes satisfy the model technical specifications for the voltage-based repair criteria as specified in GL 95-05 will ensure adequate structural and leakage l
integrity and, therefore, are acceptable.
4.0 STAFF CONCLUSION The licensee submitted an application for a license amendment to permit the use of the permanent voltage-based repair criteria for steam generator tubes i
at Farley Unit 1.
The staff has reviewed the proposed amendment and concludes that the proposed permanent alternate repair criteria are consistent with GL.
95-05 and are acceptable. The staff also concludes that adequate structural and leakage integrity can be assured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied. The staff's approval of the proposed voltage-based repair 4
criteria is based in part on the licensee being able to successfully demonstrate after each inspection outage the conditional probability of burst and the primary-to-secondary leakage during a postulated main steam line break will be acceptable in accordance with the guidance in GL 95-05. The licensee may incorporate the proposed permanent alternate repair criteria into the TS for Farley Unit 1.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amendment. The State official had no comments.
_g_
i
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the 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 individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 4353 dated January 29,1997). The amendment also changes reporting or recordkeeping requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (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 amendment.
7.0 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 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 Contributor: John Tsao Date:
March 24, 1997 4
l