ML20216C575
| ML20216C575 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/08/1998 |
| From: | Feigenbaum T NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO) |
| To: | NRC |
| Shared Package | |
| ML20216C563 | List: |
| References | |
| GL-91-04, GL-91-4, NUDOCS 9804140492 | |
| Download: ML20216C575 (7) | |
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This License Amendment Request is submitted by North Atlantic Energy Service Corporation pursuant to 10CFR50.90. The following information is enclosed in support of this License Amendment Request:
Section 1 Introduction and Safety Assessment for Proposed i
Change Section ll Markup of Proposed Change Section 111 Retype of Proposed Change Section IV Determination of Significant Hazards for Proposed Change Section V Proposed Schedule for License Amendment lasuance and Effectiveness i
Section VI EnvironmentalImpact Assessment e
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Sworn and Subscribed day of [
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>M d44 Ted C.Feigenbaum /
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V~otary Public Executive Vice President and Chief Nuclear Officer N
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Section I Introduction and Safety Assessenent for the Proposed Changes l
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L INTRODUCTION AND SAFETY ASSESSMENT OF PROPOSED CIIANGES A.
Introduction License Amendment Request (LAR 98-03) is the second submittal in a planned series of License Amendment Requests which propose changes to the Seabrook Station Technical Specifications to accommodate fuel cycles of up to 24 months for those selected surveillances that are currently performed at each 18-month or other outage interval.
The Technical Specifications proposed to be amended are:
4.4.5.3 Steam Generators - Inspection Frequencies 3.4.6.2c Reactor Coolant System Leakage 3/4.4.5 Steam Generators Bases 3/4.4.6.2 Operational Leakage Bases The proposed changes to the Seabrook Station Technical Specifications (TS) have been evaluated and modified in accordance with the generic guidance contained in NRC Generic Letter (GL) 91-04,
" Changes in Technical Specification Surveillance Intervals To Accommodate A 24-Month Fuel Cycle."
For the proposed changes contained herein, GL 91-04 requires that licensees evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. The evaluation should:
support a conclusion that the effect on safety is small, j
a confirm that historical plant maintenance and surveillance data support this conclusion and, e
1 confirm that assumptions in the plant licensing basis would not be invalidated on the basis of e
performing any surveillance at the bounding surveillance interval limit provided to accommodate a 24-month fuel cycle.
GL 91-04 further states that in consideration of these confirmations, the licensees need not quantify the cffect of the change in surveillance intervals on the availability ofindividual systems or components.
Surveillance Requirement (SR) 4.4.5.3 is currently performed at intervals of not less than 12 nor more than 24 calendar months after the pre-service inspections. If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into Category C-1 (as defined in T/S 4.4.5.2.) or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months. However, for plants having inspection resuhs in the C-2 Category from inspections of steam generators (SGs) during either of the two previous inspections, the bounding interval for the next inspection would be 24 months from the last inspection.
A 24-month inspection interval may not always coincide with the next refueling outage when operating on fuel cycles of up to 24 months, particularly if any out,ge time is accumulated over the duration of the fuel cycle or if startup for the next fuel cycle is delayed following the completion of a SG inspection.
Therefore, the proposed changes to Surveillance Requirement 4.4.5.3 and Limiting Condition for Operation 3.4.6.2.c of the Seabrook Station Technical Specifications provide an alternative to compensate for any delay that could cause the interval for steam generator inspections to occur near the end of a 24-month fuel cycle but before the refueling outage. The alternative includes,(1) an increase in the sample size of tubes examined. (2) a suitable analysis of the integrity of steam generator tubes, if the Page 2
inspectioh results are in a C-2 or a C-3 category, and (3) for plant operation beyond 24 months from the previous steam generator tube inspection when the results of either of the two previous inspections are in
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the C-2 category, the reactor coolant system leakage through any one steam generator shall not exceed 100 gallons per day.
j The proposed changes will modify existing Surveillance Requirement 4.4.5.3.a to reflect the increase in sample size of tubes examined and the requirement of performing an engineering assesament of the j
steam generator tubes if the inspection results are in a C-2 or C-3 Category. Surveillance Requirement j
4.4.5.3.b will be modified to reflect the interval of steam generator tube inspections of either 30 or 40 i
months depending on the results of the two consecutive inspections. Surveillance Requirement 4.4.5.3.d j
has been added to clarify that the provisions of Technical Specification Section 4.0.2 do not apply to the extended steam generator inspection interval because Technical Specification Section 4.4.5.3.a addresses those conditions under which the 24-month surveillance interval for steam generator tube inspections may be extended. These proposed changes a SR 4.4.5.3 are consistent with GL 91-04 suggested Technical Specification wording.
Bases Section 3/4.4.4.5," Steam Generators," has been modified to reflect the intent of the engineering assessment for steam generator integrity addressed in Technical Specification Section 4.4.5.3.a. This addition addresses those conditions under which the 24-month surveillance interval for steam generator tube inspections may be extended. This proposed change to Bases 3/4.4.4.5 is consistent with GL-91-04 suggested Technical Specification wording.
l The proposed change to Technical Specification Limiting Condition for Operation 3.4.6.2.c will add a more restrictive limit of Reactor Coolant System leakage through the steam generators (100 gallons per day for any steam generator) for the Category C-2 condition with steam generator tube inspections beyond 24 months. This proposed change to TS 3.4.6.2.c is consistent with GL 91-04 suggested Technical Specification wording. In addition, the associated Bases Section, 3/4.4.6.2, " Operational Leakage" has been modified to reflect the more restrictive limit being imposed for Reactor Coolant System leakage through steam generators, for steam generators with Category C-2 tube inspection results with inspection intervals beyond 24 months.
In summary, the proposed changes are consistent with the suggested changes specified in GL 91-04. The technical evaluation of the components surveilled by the TS surveillance requirements addressed herein conclude that the effect on plant safety by the proposed extension at the bounding surveillance interval of 30 months to be insignificar* Ealysis of historical surveillance data indicates tube degradation of the type experienced at Seabros...ution, Anti-Vibration Bar (AVB) Wear, will not reduce the margins of safety required by Regulatory Guide 1.121 for fuel cycles extended up to 24 months. The proposed changes do not alter the intent or method by which the surveillances are conducted, do not involve any physical changes to the plant, do not alter the way any structure, system or component (St C) functions, and do not modify the manner in which the plant is operated. As such, the proposed chanh.es to extend the surveillance intervals will not degrade the ability of any SSC to perform its safety function. In addition, the performance of the referenced surveillances at the bounding surveillance interval of 30 months (24 months plus 25% extension) does not adversely cffect nor invalidate assumptions in the plant licensing basis.
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Safety Assessment of Proposed Changes i
l-There are four steam generators in the Reactor Coolant System (RCS), one per loop. The function of the f
steam generators is to remove the heat from the reactor coolant system in order to produce high quality steam to drive the turbine generator.
The four Westinghouse Model F steam generators are original' components designed to ASME Section 111. Each steam generator contains 5626 Thermally Treated (TT), Inconel 600 U-tubes (SB-163),
hydraulically expanded into the tubesheet at each end. The steam generator tubing is nominally 0.688" i
O.D. with a 0.040" wall thickness. The tube bundle is supported with a series of "V" shaped Anti-l Vibration Bars (AVBs) in the U-bend region and eight stainless steel Tube Support Plates (TSPs), which j
includes the flow distribution baffic as TSP #1 in the straight leg regions.
t The tubesheet is drilled on a square pitch with a 0.98" spacing. Each tube is identified by a row and colman number. Rows are orientated parallel to the primary side head divider plate whereas the columns l
are perpendicular to the divider plate.
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural i
integrity of this portion of the RCS will be maintained, since the tubes comprise a significant fraction of the surface area of the reactor coolant pressure boundary (RCPB). The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice l
inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
As an adjunct to the inservice inspection program, operational limits are imposed by the Technical Specifications for steam generator primary to secondary leakage to ensure 1) that the dosage contribution from the tube leakage will be limited te a small fraction of 10 CFR Part 100 dose guideline values in the l
event of either a steam generator tube rupture or steam line break, and 2) that tube integrity is maintained l
in the event of a steam line rupture or under LOCA conditions.
l A key component of inservice inspection of steam generator tubing is the use of eddy current testing (ECT) techniques in locating defect areas in the steam generator tubing and for assessing the overall condition of the tubing. Tubes with imperfections that exceed the minimum acceptable tube wall thickness and operational limit are removed from service by plugging techniques.
Throughout Seabrook Station's operating history, steam generator tube plugging has been primarily due to AVB wear. AVB wear is a degradation process of steam generator tubes due to mechanical rubbing of the steam generator tubes with the anti-vibration bars. This degradation process is caused by flow-induced vibration of the steam generator tubes.
The steam generators have been inspected as groups of two steam generators (RC-E-II A & RC E-llD or RC-E-llB & RC-E-IIC), except for the first refueling outage (OR01) where all four steam generators were inspected. During OR01,30% of the total number of steam generator (RC-E-ll A, RC-E-llB, RC-E-lIC, & RC-E-11D) tubes were inspected. Subsequently, the inspection sample was increased to 40%
of the total number of tubes (RC E-1IB & RC-E-llc or RC-E-ll A & RC-D-llD depending on which 8
outage) for the second refueling outage (OR02) and third refueling outage (OR03). During the fourth refueling outage (OR04),100% of tubes which were suspected as having the potential of experiencing AVB wear (43% of the total for RC-E-II A & RC-E-ilD) were inspected. During the fifth refueling outage (OROS),100% of the total number of tubes for RC-E-l1B & RC-E-11C were inspected.
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The OROS inspection identified AVB wear in regions of the steam generators which previously were thought as not having the potential of AVB wear, i.e., tubes contained in the first 24 rows or less, closest to the divider plate. Seven (7) tubes were required to be plugged due to AVB wear. This brings the total number of steam generator tubes plugged due to AVB wear to 24 tubes. No other forms of inservice tube degradation have been identified after 5 cycles of operation other than loose part wear.
Loose part wear is identified by eddy current testing and Foreign Object Search and Retrieval (FOSAR) activities. An indication of a possible loose part requires an assessment of the indication with resolutions such as tube plugging, part retrieval and/or engineering evaluation.
Observed Seabrook tube AVB wear rates were used to_ evaluate the proposed extended inspection intervals against an allowable 75% through-wall structural limit, as specified in Regulatory Guide 1.121,
" Bases For Plugging Degraded PWR Steam Generatoi Tubes." The analysis developed several cases to project AVB flaw growth rate through future cycles. In each case evaluated, it has been determined that the allowable 75% through-wall structural limit would not be exceeded.
During the most recent outage (OR05, May / June 1997), steam generator eddy current inspection identified a total of 163 tubes (434 Haws) from steam generators B and C as containing AVB Haws.
Steam generators A and D were inspected during 1995 (OR04), in which a total of 161 tubes (378 flaws) were identified containing AVB flaws. A comparison of these eday current inspections with previous inspections (S/G A & D - OR02,1992; S/G B & C - OR03,1994) was completed to determine the AVB wear rates. The average flaw growth rate for each time period between inspections was determined (22%/882 EFPDs for S/G A & D, and 15%/942 EFPDs for S/G B & C). The larger AVB wear flaw growth rate,(22%/882 EFPDs for S/G A & D) was used to show in the analysis that Seabrook's steam generator tubes will not exceed the 75% throughwall structural limit due to AVB flaw growth.
Wear rate and structural analyses for another Model F steam generator have been performed. A comparison of Seabrook Station's steam generator AVB flaw progression rates to these analyses shows the Seabrook Station steam generator AVB flaw progression rates to be considerably less. The Seabrook Station analyses concluded that a structural limit of 75% throughwall can be assumed for AVB wear in i
Seabrook Station steam generator tubing based on Regulatory Guide 1.121 criteria.
l Therefore, given the present AVB growth rate at Seabrook Station, the steam generator tube inspection schedule, and the projected fuel cycle length, the maximum projected AVB flaw depth will not exceed l
the 75% throughwall structural limit at Seabrook Station.
Technical Specification Bases 3/4.4.6.2, Operational Leakage, states that the total steam generator tube leakage limit of I gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The I gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a steam line rupture or under LOCA conditions.
1 To provide additional margin to accommodate a tube flaw which might grow at a c4 - :r than expected rate, for steam generators with Category C-2 tube inspection results with inspectio.
mrvals beyond 24 months, a more restrictive operational leakage limit of 100 gallons per day per steam generator is being proposed to TS 3.4.6.2 and its associated Bases. The revised limit is intended to provide additional assurance that should a significant leak be experienced in service the plant will be shut down in a timely i
manner. Furthermore, the revised limit is consistent with GL 91-04 suggested changes.
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Other Potential Degradation Mechanisens f
As stated previously, the Seabrook Station steam generator inspections over the past five cycles of operation have not identified secondary side flaws or tube degradation mechanisms other han AVB wear l
and loose parts wear. However, as is the case with any steam generator, the potential exists that a previously non-existent damage mechanism could develop in the future, such as degradation mechanisms associated with secondary and primary water chemistry.
North Atlantic has demonstrated the ability to effectively control secondary chemistry to industry recognized standards and to pro-actively address potential future issues such as controlling ' steam generator tube fouling by chemistry means in an integrated and programmatic manner. As such, secondary side steam generator corrosion is not anticipated to be a significant issue for the foreseeable future (i.e., within the next 10 years).
l The Seabrook Station steam generators are considered to be less susceptible to Primary Water Stress i
Corrosion Cracking (PWSCC) because they are fabricated with thermally treated (TT) Alloy 600 tubing.
l In addition, the U-bends of the lower 10 rows of tubes have been stress relieved. Based on current industry data, it is concluded that TF Alloy 600, in combination with lower ro v stress relief is expected to provide better resistance to PWSCC within such stressed areas as tube / tubesheet roll transitions and U-bend areas.
The steam generator inspection programs will continue to use appropriate non-destructive examination (NDE) techniques to effectively monitor for the potential development of steam generator secondary side tube flaws and degradation mechanisms and will continue to include the appropriate NDE techniques for j
PWSCC detection. This inspection program also addresses the requirements of USNRC Generic Letter
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95-03, "Circumferential Cracking of Steam Generator Tubes," for detection of circumferential cracking.
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During the November,1995 (OR04) steam generator inspection, a sample of tubes using a probe qualified for circumferential crack detection per Appendix H of the EPRI Steam Generator Examination Guidelines revealed no indications of degradation.
l In conclusion, the effect on plant safety by the proposed extension of Steam Generator tube inspections I
at the bounding surveillance interval of 30 months to support fuel cycles of up to 24 months has been determined to be insignificant. Analysis of historical surveillance data indicates tube degradation of the type experienced at Seabrook Station, i.e., AVB Wear, will not reduce the margins of safety required by Regulatory Guide 1.121 for fuel cycles extended up to 24 months. The proposed changes do not alter the I
intent or method by which the surveillances are conducted, do not involve physical changes to the plant, do not alter the way a structure, system or component (SSC) functions, and do not modify the manner in which the plant is operated. As such, the proposed changes to extend the surveillance intervals will not degrade the ability of a SSC to perform its safety function. In addition, the performance of the referenced surveillances at the bounding surveillance interval of 30 months (24 months plus 25% extension) does not adversely affect nor invalidate assumptions in the plant licensing basis.
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