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{{Adams
#REDIRECT [[L-2014-252, In-Service Inspection Plans Fourth Ten-Year Interval Relief Request 2]]
| number = ML14224A010
| issue date = 08/01/2014
| title = St. Lucie, Unit 2, In-Service Inspection Plans Fourth Ten-Year Interval Relief Request 2
| author name = Catron S
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000389
| license number =
| contact person =
| case reference number = L-2014-252
| document type = Letter
| page count = 20
| project =
| stage = Other
}}
 
=Text=
{{#Wiki_filter:0FPL.August 1, 2014L-2014-252 10 CFR 50.410 CFR 50.55aU. S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, DC 20555Re: St. Lucie Unit 2Docket Nos. 50-3 89In-Service Inspection PlansFourth Ten-Year IntervalUnit 2 Relief Request 2Pursuant to 10 CFR 50.55a(a)(3)(ii)
FPL requests an alternative to the requirements of ASMEBoiler & Pressure Vessel Code, Section XI, paragraph IWB-3132.2, "Acceptance byRepair/Replacement Activity."
The original alloy 600 small bore nozzles and pressurizer heatersleeves in the St. Lucie Unit 2 reactor coolant system (RCS) have been replaced with alloy 690nozzles and heater sleeves.
The nozzle welds and pressurizer heater sleeves have been repairedusing the "half-nozzle" technique or the "sleeve" technique.
The bases and justification for the"half-nozzle" and "sleeve" repair techniques are within the Attachment to this letter.Please contact Ken Frehafer at 772-467-7748 if there are any questions about this submittal.
Sincerely,
ýtefi~e C tronLicen ng ManagerSt. Lucie PlantAttachment SC/KWFcc: USNRC Regional Administrator, Region IIUSNRC Senior Resident Inspector, St. Lucie Units I and 2Florida Power &Light Company6501 S. Ocean Drive, Jensen Beach, FL 34957 L-2014-252 Attachment Page 1 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(ii)
--Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety--1 ASME Code Component(s)
AffectedSmall bore alloy 600 nozzles welded to the reactor coolant piping hot legs andpressurizer and alloy 600 heater sleeves welded to the pressurizer St. Lucie (PSL) Unit 2Reactor Coolant Piping Nozzle DetailsFPL Drawing Numbers:
2998-18705 Rev. 2, 2998-18706 Rev. 2Pressurizer Nozzle DetailsFPL Drawing Numbers:
2998-19321 Rev. 0, 2998-19466 Rev. 0, 2998-19467 Rev. 0Pressurizer Heater SleevesFPL Drawing Numbers:
2998-16985 Rev. 52. Applicable Code Edition and AddendaThe Code of record for St. Lucie Unit 2 (PSL-2) is the 2007 Edition with 2008Addenda of ASME Boiler and Pressure Vessel Code, Section Xl, "Rules forInservice Inspection of Nuclear Power Plant Components."
: 3. Applicable Code Requirement Pursuant to 10 CFR 50.55a (a)(3)(ii)
FPL requests an alternative to therequirements of ASME Boiler & Pressure Vessel Code, Section XI, paragraph IWB-3132.2 "Acceptance by Repair/Replacement Activity.
" A component whosevolumetric or surface examination detects flaws that exceed the acceptance standards of Table IWB-3410-1 is unacceptable for continued service until theadditional examination requirements of IWB-2430 are satisfied and thecomponent is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWB-3000.
"
L-2014-252 Attachment Page 2 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0FPL requests an alternative to the requirements of ASME Boiler & PressureVessel Code, Section XI, IWB-3132.2 and the repairs take no other exceptions toapplicable ASME Code requirements.
: 4. Reason for RequestSmall bore nozzles were welded to the interior of the hot leg of the reactorcoolant piping and pressurizer and heater sleeves were welded to the interior ofthe pressurizer during original fabrication of the piping and pressurizer.
Industryexperience has shown that cracks may develop in the alloy 600 nozzle basemetal, heater sleeve base material, or in the weld metal joining the nozzles orheater sleeves to the reactor coolant pipe or pressurizer, resulting in leakage ofthe reactor coolant.
The cracks are believed to be caused by primary waterstress corrosion cracking (PWSCC).
The potential flaws, through the weld, basematerial, or both, cannot be determined.
To remove all possible potential flaws requires accessing the internal surface ofthe component to grind out the attachment weld and any remaining nozzle basemetal. Such an activity would result in high radiation exposure to the personnel
: involved, which is considered a hardship.
Grinding within the components alsoexposes personnel to safety hazards.
Additionally, grinding inside the reactorcoolant piping or pressurizer increases the possibility for the introduction offoreign material that could damage the fuel cladding.
The NRC approved topicalreport (TR) in WCAP-15973-P-A, Rev 0 "Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Programs"
[1], and the following
: section, "Proposed Alternative and Basis forUse", support the conclusion that there is no compensating increase in the levelof quality or safety as a result of removal of the flawed metal.5. Proposed Alternative and Basis for UseProposed Alternative The alloy 600 small bore nozzles and pressurizer heater sleeves in the PSL-2Reactor Coolant System (RCS) have been replaced with alloy 690 nozzles andheater sleeves.
The nozzle welds and Pressurizer heater sleeves have beenrepaired using the "half-nozzle" technique or the "sleeve" technique.
The originalnozzles and heater sleeves were repaired by relocating the attachment weld fromthe inside surface of the pipe or pressurizer to the outside surface of the pipe orpressurizer.
The alloy 600 small bore nozzle repairs at PSL-2 are shown inTable 1. Note that twenty-two (22) of the twenty-six (26) nozzle replacements L-2014-252 Attachment Page 3 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0and all heater sleeve replacements were performed preventatively without thepresence of a flaw or leak.In the "half-nozzle" technique, Figure 1, design A and B, the components are cutoutboard of the partial penetration weld between the nozzles and pipe orpressurizer wall for the nozzles or between the heater sleeves and pressurizer wall for the heater sleeves, approximately midwall.
The cut sections of the alloy600 nozzles and heater sleeves are replaced with short sections (half-nozzles) ofalloy 690, which are then welded to the outside surfaces of the pipe orpressurizer.
The remainders of the alloy 600 nozzles and heater sleeves,including the partial penetration welds, remain in place without correction.
In the "sleeve" technique, Figure 1, design C and D, the entire nozzle is removedby machining and the bore diameter is slightly enlarged.
Subsequently an alloy690 sleeve is inserted into the bore and rolled into place. The end of the sleeveat the interior surface of the piping or the pressurizer is either roll expanded orwelded to the interior surface of the piping or pressurizer, essentially eliminating corrosion of the carbon steel by stopping the replenishment of borated solution incontact with the carbon steel. An alloy 690 nozzle is inserted into the sleeve andthe nozzle and sleeve are welded to the exterior of the piping or pressurizer.
The weld joint configurations shown in Figure 1 are illustrative only. Twelve (12)of the twenty-six (26) repaired alloy 600 nozzles and all of the pressurizer heatersleeves are welded to pads deposited on the exterior surface of the pressurizer or piping using a temper bead technique or directly to the piping surface.The remnant material (weld metal, nozzles and heater sleeves) will not receiveadditional examination.
The new pressure boundary welds located on theexterior surface of the piping or pressurizer will be examined in accordance withthe applicable requirements of the ASME Boiler and Pressure Vessel CodeSections III and XI.Basis For UseSection 2.3 of the TR in Reference 1 evaluates the effect of component corrosion resulting from primary coolant in the "half-nozzle" crevice region between theremnant alloy 600 nozzles and replacement alloy 690 nozzle. In addition, Section 2.5 of the TR in Reference 1 evaluates the effect of component corrosion resulting from primary coolant in a confined
: crevice, like the "sleeve" repair,where the volume of the solution is such that the solution cannot be replenished.
In the "half-nozzle" repair, a small gap remains between the remnant of theoriginal alloy 600 component and the new alloy 690 component.
As a result, L-2014-252 Attachment Page 4 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0primary coolant (borated water) will fill the crevice between the nozzle or heatersleeve and the pipe or the pressurizer wall. Low alloy and carbon steels used forreactor coolant systems components are clad with stainless steel to minimizecorrosion resulting from the exposure to borated primary coolant.
Since acrevice exists, the low alloy and carbon steels are exposed to borated water.Therefore, the corrosion rates addressed in the "half-nozzle" repair is based onthe corrosion analysis in Section 2.3 of the TR of Reference 1.The "sleeve" repair was not specifically evaluated in the TR of Reference 1.However, Section 2.5 of the TR [1] provides an alternate estimate of carbon andlow alloy steel corrosion.
The corrosion rate previously described is applicable tothe carbon and low alloy steel exposed to bulk solutions of boric acid and not tosolutions confined in a crevice where the volume of the solution is such that thesolution cannot be replenished or refreshed.
The geometry of the "sleeve" repairresults in a tight crevice between the alloy 690 sleeve and the base metal of thehot leg piping or the pressurizer, which is equivalent or even tighter than thecrevice evaluated in Section 2.5 of the TR [1]. Therefore, the corrosion rateshown in Section 2.5 of the TR in Reference 1 is used to evaluate the "sleeve"repair.Reference 1, demonstrates that the carbon and low alloy steel RCS components at PSL-2 will not be unacceptably degraded by general corrosion as a result ofthe implementation of replacement of small diameter alloy 600 nozzles andheater sleeves.
Although some minor corrosion may occur in the crevice regionof the replaced nozzles and heater sleeves, the degradation will not proceed tothe point where ASME Boiler & Pressure Vessel Code requirements will beexceeded before the end of plant life, including the period of extended operation.
: Further, available laboratory data and field experience indicate that continued propagation of cracks into the carbon and low alloy steels by a stress corrosion mechanism is unlikely.
Additionally, Reference 1 evaluates the effects of propagation of the flaws, left inplace from the previous nozzles and welds, by fatigue crack growth and stresscorrosion cracking mechanisms.
Postulated flaws were assessed for flaw growthand flaw stability as specified in the ASME Boiler & Pressure Vessel Code,Section Xl and the results demonstrate compliance with the requirements of theASME Boiler & Pressure Vessel Code, Section X1.Reference 2 (NRC letter dated January 12, 2005, Final Safety Evaluation forTopical Report WCAP-15973-P, Rev 1) states "The staff has found that WCAP15973-P, Revision 01, is acceptable for referencing in licensing applications forCombustion Engineering designed pressurized water reactor to the extent L-2014-252 Attachment Page 5 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0specified and under the limitations delineated in the TR (Topical Report) and inthe enclosed SE (Safety Evaluation)".
Sections 4.1, 4.2, and 4.3 of the Safety Evaluation (SE) in Reference 1 presentadditional conditions to assess the applicability of the TR in Reference
: 1. TheFPL assessment for each additional condition is provided below. The FPLassessment is in italic font. The discussion shows that Reference 1 is applicable to PSL-2.A. Section 4.1 of the SE in Reference 1 states that "Licensees seeking to usethe methods of the TR will need to perform the following plant-specific calculation in order to confirm that the ferritic portions of the vessels or pipingwithin the scope of the TR will be acceptable for service through the licensedlives of their plants (40 years if the normal licensing basis plant life is used or60 years if the facility is expected to be approved for extension of theoperating license):"
: 1. "Calculate the minimum acceptable wall thinning thickness for theferritic vessel or piping that will adjoin to the MNSA repair or half-nozzle repair."FPL Assessment:
Based on Item 4 in Reference 3, the corrosion calculations herein will address the Limiting Allowable
: Diameter, asdescribed in Reference 4, in lieu of the minimum acceptable wallthickness for the vessel or piping. The Limiting Allowable Diameters, as described in Reference 4, for the various nozzles under evaluation are shown in Tables 2A and 2B and the associated weld jointconfigurations are shown in Figure 1 herein.2. "Calculate the overall general corrosion rate for the ferritic materials based on the calculation methods in the TR, the general corrosion rates listed in the TR for normal operations, startup conditions (including hot standby condition) and cold shutdown conditions and therespective plant-specific times in (in-percentages of total plant life) ateach of the operating modes."FPL Assessment.
The overall general corrosion rate was determined using the calculation methods in Section 2.3 of the TR in Reference Iand PSL-2 generation data from 1/1/1995 to 2/28/2014.
Thepercentage of total plant time spent at each of the temperature conditions follows:High temperature conditions 90.5%
L-2014-252 Attachment Page 6 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0Intermediate temperature conditions 2.0%Low temperature conditions 7.5%The corrosion rate for each temperature condition is taken from the TR[1] and is shown as follows:High temperature conditions 0.4 mpyIntermediate temperature conditions 19.0 mpyLow temperature conditions 8.0 mpyThe overall corrosion rate was determined using the above time attemperature data, corrosion rate at temperature data, and formula 1 ofthe TR [1] as follows:CR = 0.90.5 X 0.4 mpy + 0.02 X 19 mpy + 0.075 X 8 mpyResulting in an overall corrosion rate of 1.34 mpy. This corrosion rateis applicable only to the" half-nozzle" repair.The overall general corrosion rate for the "sleeve" repair is based onSection 2.5 of the TR in Reference'l, which addresses corrosion occurring in a tight crevice and describes the mechanism thatdifferentiate crevice corrosion from corrosion occurring in the bulk fluidenvironment.
The corrosion rate discussed in Section 2.5 of the TR [1]is applicable to any tight crevice geometry within the bounds of theevaluation of Section 2.5 of the TR [1]. The geometry of the "sleeve"repair results in a tight crevice between the alloy 690 sleeve and thebase metal of the hot leg piping or pressurizer, which is equivalent oreven tighter than the crevice evaluated in Section 2.5 of the TR [1].Therefore, the overall corrosion rate for the "sleeve" repair is boundedby the corrosion rate discussed in Section 2.5 of the TR in Reference 1.3. "Track the time at cold shutdown conditions to determine whether thistime does not exceed the assumptions made in the analysis.
If theseassumptions are exceeded, the licensees shall provide a revisedanalysis to the NRC and provide a discussion on whether volumetric inspection of the area is required."
FPL Assessment:
In accordance with section 2.3.4 of the SE inReference 1, the corrosion rate for CE plants is based on a time split of88 percent at operating conditions, 2 percent at intermediate L-2014-252 Attachment Page 7 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0temperature startup conditions, and 10 percent at low temperature outage conditions.
An assessment of operating data for PSL-2 from1/1/1995 through 2/28/2014 shows a time split of 90.5 percent atoperating conditions, 2.0 percent at intermediate temperature startupconditions, and 7.5 percent of plant time at low temperature outageconditions.
Therefore, the time at cold shutdown does not exceed theassumptions made in the analysis.
The plant operating conditions will be reassessed for the resubmittal ofthis relief request at the start of the next inspection
: interval, whichbegins in August 2023. There is no need to track plant operating conditions during the remainder of the current inspection
: interval, asthere is sufficient wall thickness in the more limiting hot leg piping tomaintain the limiting allowable diameter until this reassessment ismade. As shown in the TR of Reference 1, the most severe corrosion rate for steady state conditions,
: i. e. at power or shutdown, wouldoccur during outage or shutdown conditions with a corrosion rate of 8mpy. Using the calculated corrosion rate of 1.34 mpy, from 2013 forone year, the wall would have experienced a radial loss of 0.001 in. todate. If the plant remained shut down for the remainder of theinspection
: interval, approximately 9 years, and experienced corrosion of the steel at the rate shown in the TR [1], approximately 8 mpy, therewould be an additional loss of 0.072 in. of wall thickness.
The totalloss, 0.001 in. plus 0.072 in., would equal 0.073 in. Doubling the lossto account for a diametrical change and adding the diameter of 1.063in. from Table 2A results in a diameter of 1.210 in. at the start of thenext inspection interval.
A diameter of 1.2 10 in. is less than the limitingdiameter of 1.270 in. identified in Reference 12 of WCAP-15739-P-A, Rev. 0. This calculation was performed for a "half-nozzle" repair only.As shown below the corrosion rate for the "sleeve" repair has a lifetimediametrical loss of 0.025 in. and therefore is bounded by thecalculation for the "half-nozzle" repair.4. "Calculate the amount of general corrosion based thinning for thevessels or piping over the life of the plant, as based on the overallgeneral corrosion rate calculated in Step 2 and the thickness of theferritic vessel or piping that will adjoin to the MNSA repair or half-nozzle repair."FPL Assessment:
The amount of corrosion will be determined for twocases; 1) the overall general corrosion rate that is applicable to the"half-nozzle" repairs and 2) the corrosion rate for tight crevices that isapplicable to the "sleeve" repairs.
L-2014-252 Attachment Page 8 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0Table I shows the first "half-nozzle" repair and first "sleeve" repair tothe piping and to the pressurizer.
For the piping, the first "half-nozzle" repair was made in 2003 and the first "sleeve" repair was made in1989. For the pressurizer, the first "half-nozzle" repair was made in1994, and the first "sleeve" repair was made in 1995. The pressurizer heater sleeves used the "half-nozzle" repair in 2011.The plant license was renewed and it expires on April 6, 2043. Thefirst "half-nozzle"
: repairs, made in 1994, can expect to see 49 moreyears of service from the year of the repair to the year the plant licenseexpires.
Applying the "half-nozzle" corrosion rate from step 2, 1.34mils per year, for 49 years results in a radial material loss of 65.6 mils(diametrical loss of 131 mils) for the "half-nozzle" repairs.
"Thematerial loss at the end of the plant life for "half-nozzle" repairsperformed after 1994 are bounded by the calculated material loss forthe first "half-nozzle" repair because repairs made after 1994 will haveless years of service than the first "half-nozzle repair".For the pressurizer heater sleeves repair, the "half-nozzle" repairsmade in 2011 can expect to see 32 more years of service from theyear of the repair to the year the plant license expires.
Applying the"half-nozzle" corrosion rate from step 2, 1.34 mils per year, for 32years results in a radial material loss of 42.9 mils (diametrical loss of86 mils) for the heater sleeve repairs.
Therefore, the pressurizer heater sleeve repairs are bounded by the calculated material loss forthe first "half-nozzle" repair because the heater sleeve repairs will haveless years of service than the first "half-nozzle repair".The first "sleeve" repairs were made in 1989 and can expect to see 54more years of service from the year of the repair to the year the plantlicense expires.
As shown in Section 2.5 of the TR in Reference 1, areasonable estimate of the lifetime corrosion resulting from a tightcrevice will be a radial material loss of 12.5 mils (diametrical loss of 25mils) which is considered applicable to the "sleeve" repairs.5. "Determine whether the vessel or piping is acceptable over theremaining life of the plant by comparing the worst case remaining wallthickness to the minimum acceptable wall thickness for the vessel orpipe."FPL Assessment:
In Tables 2A and 2B, the third column from the leftlists the nozzle bore in the piping or pressurizer resulting from the L-2014-252 Attachment Page 9 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0replacement of the Alloy 600 nozzles and heater sleeves.
Also inTables 2A and 2B, the radial material loss, from Step 4 above, isdoubled and added to the repair bore diameter.
The resultant nozzlerepair bore diameter is compared to the Limiting Allowable
: Diameter, from Step 1. For the nozzle locations shown, the resultant diameter isless than the Limiting Allowable Diameter.
Therefore, the hot leg piping and the pressurizer are acceptable for theremaining life of the plant.B. Section 4.2 of the SE in Reference 1 states that "Licensees seeking toreference this TR for future licensing applications need to demonstrate that:"1. "The geometry of the leaking penetration is bounded by thecorresponding penetration reported in Calculation Report CN-CI-02-71, Revision 01."FPL Assessment:
Plant specific calculations to evaluate fatigue crackgrowth associated with small diameter nozzles have been performed and are reported in Reference
: 5. The calculations and results areequivalent to Calculation Report CN-CI-02-71, Revision
: 01. Thecalculations of Reference 5 do not address the pressurizer heatersleeves.
: However, the geometry of the PSL-2 pressurizer heatersleeves is equivalent to that shown in Calculation Report CN-CI-02-71, Rev. 1. Therefore, the geometry of the nozzles on PSL-2 are boundedby Calculation Report CN-CI-02-71, Rev. 1. Reference 5 wassubmitted to the NRC as part of the St. Lucie License Renewal activity, which resulted in an extended license for PSL-2.2. "The plant-specific pressure and temperature profiles in the pressurizer water space for the limiting curves (cooldown curves) do not exceedthe analyzed profile shown in Figure 6-2 of Calculation Report CN-CI-02-71, Revision 01, as stated in Section 3.2.2 of this SE."FPL Assessment:
The TR in Reference I indicates that the pressurizer cool down profile analyzed is a 200 degree F per hour cooldown ratefrom 653 degrees F to 200 degrees F followed by a 75 degree F perhour rate to 120 degrees F. The TR [1] indicates that the fatigueevaluation results are not affected by the choice of cooldown rate from653 degrees F to 200 degrees F and that the only concern is when themetal temperature is less than 200 degrees F, which is when thematerial toughness begins to significantly decrease.
L-2014-252 Attachment Page 10 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0Cooldown of the pressurizer water space is administratively controlled by a plant procedure to a maximum rate of 75 degrees F per hour fornormal operation, which is within the rates shown in Figure 6-2 of CN-CI-02-71.
Additionally, the pressurizer water temperature is recordedduring cooldown until the water temperature is less than 120 degreesF.Therefore, the temperature profile in the pressurizer water space doesnot exceed the analyzed profile shown in Figure 6-2 of CN-CI-02-71.
: 3. "The plant-specific Charpy USE data shows a USE value of at least 70ft-lb to bound the USE value used in the analysis.
If the plant-specific Charpy USE data does not exist and the licensee plans to use CharpyUSE data from other plants pressurizers and hot leg piping, thenjustification (e.g., based on statistical or lower bound analysis) has tobe provided."
FPL Assessment:
Charpy USE value of 70 ft-lb was used to support anEPFM analysis of the pressurizer lower shell and the pressurizer lowerhead. The analysis was not performed on the upper head because theupper head is not affected by the large in-surge transient or thermalstress that occurs at the lower head and lower shell. When thepressurizer was built, Charpy USE data for the pressurizer was notrequired and was not determined.
The Charpy impact data for the two(2) lower shell plates, the upper head, and the bottom head of thepressurizer is summarized in Table 3. The summarized data is theaverage of the impact test data reported in the materials certification reports.The Charpy impact data and USE for six (6) plates in the reactorvessel (RV) shell is summarized in Table 3. The summarized data isthe average of the impact test data included in the material certification reports for the RV plates and were chosen at temperature and shearlevels comparable to those used for testing the pressurizer materials.
The two (2) lower shell plates, the upper head, and the bottom head ofthe pressurizer are similar to the six (6) RV plates. All ten (10) itemswere made to the same alloy specification, SA-533 Gr. B Cl. 1, havesimilar chemistry, and received similar heat treatment.
Lukens Steelsupplied the pressurizer upper and bottom heads, and the six (6) RVplates. Marrel Freres supplied the two (2) pressurizer lower shellsplates. Since the ten (10) items are similar, it can be reasonably expected that the USE data for the two (2) lower shell plates, the upper L-2014-252 Attachment Page 11 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0head, and the bottom head of the pressurizer should be comparable tothat of the RV plates, as discussed below.From Table 3, the pressurizer lower shell plate, Heat No. NR 60 466-2,exhibited an absorbed energy of 72 ft-lb and 35% shear at a testingtemperature of +20 degrees F. The USE value is the absorbed energyat 100% shear and this shear state is obtained by testing atprogressively higher temperatures.
As the testing temperature isincreased, the absorbed energy increases and the percent shearincreases.
Since this material already exhibits the required 70 ft-lb atlow temperatures, it will continue to exhibit and exceed the requiredvalue of 70 ft-lb while approaching full shear.Similarly for the pressurizer bottom head, Heat No. C4754-3, andupper head, B8618-2, the absorbed energy at +70 degrees F is 69 ft-lband the absorbed energy will increase as 100% shear is obtained.
Itcan be reasonably expected that these materials will exhibit an USE ofat least 70 ft-lb.The pressurizer lower shell plate, Heat No. NR 61 734-1, exhibited absorbed energy and % shear comparable to that of the six (6) RVplates at a testing temperature lower than the testing temperature ofthe RV plates. From Table 3, the testing temperature for Heat No. NR61 734-1 is +30 degrees F and for the RV plates is +60 degree F.Since all seven items have similar chemistry, experienced similar heattreatment, and the pressurizer lower shell exhibited impact properties similar to the RV plates at a lower testing temperature, it is reasonable to expect the USE of the pressurizer lower shell plate, Heat No. NR 61734-1, to be comparable to that of the RV plates, which exhibit USEwell in excess of 70 ft-lb.Therefore, it is reasonable to expect that the two (2) lower shell plates,the upper head, and the bottom head of the pressurizer would exhibitUSE well in excess of 70 ft-lb and that PSL-2 is bounded by theanalysis.
C. The concluding requirement of section 4.2 of the SE in Reference 1 states,"Based on the above evaluation, the staff has determined that the crack canbe left in the J-groove weld at small-bore locations for a plant life of 40 years.However, if the licensee plans on using this alternative beyond the 40 yearsand through the license renewal period, the thermal fatigue crack growthanalysis shall be re-evaluated to include the extended period, as applicable, L-2014-252 Attachment Page 12 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0and submitted as a time limited aging analysis in their license renewalapplication as required by 10 CFR 54.21(c)(1)."
FPL Assessment:
As stated above, in response to 4.1.4 of the SE inReference 1, the first small bore alloy 600 nozzle repair can beexpected to see 54 more years of service, which extends beyond theoriginal plant life of 40 years and into the license renewal period. TheSt. Lucie plant has received an extended license for both Units 1 and2. Chapter 18 of the FSAR for Unit 2, Reference 6, describes theaging management programs and time limited aging analysis activities for license renewal.
Section 18.3.7 in Chapter 18 of the FSAR [6],specifically addresses alloy 600 instrument nozzle repairs.
Thissection concludes "The flaw growth analysis of the Unit 2 pressurizer steam space alloy 600 instrument nozzle repairs has been evaluated and determined to remain valid for the period of extended operation, inaccordance with 10 CFR 54.21(c)(1)(i)."
D. Section 4.3 of the SE in Reference 1 states that "Licensees seeking toimplement MNSA repairs or half-nozzle replacements may use the WOG'sstress corrosion assessment as the bases for concluding that existing flaws inthe weld metal will not grow by stress corrosion if they meet the following conditions:"
: 1. "Conduct appropriate plant chemistry reviews and demonstrate that asufficient level of hydrogen overpressure has been implemented for theRCS and that the contaminant concentrations in the reactor coolanthave been typically maintained at levels below 10 ppb for dissolved oxygen, 150 ppb for halide ions and 150 ppb for sulfate ions."FPL Assessment:
PSL-2 follows the Reactor Coolant System (RCS)chemistry practices and limits recommended under the EPRI PWRPrimary Chemistry Guidelines for shutdown and operation conditions with no program exceptions noted in the station's Primary Chemistry Strategic plan. Chemistry data was reviewed for the period from June2005 to March 2014.Chemistry procedures state that hydrogen levels shall be >15 cc/kg butmay be < 25 cc/kg for up to 24 hours after reaching reactor criticalwithout instituting Action Level 1. Hydrogen levels may be reduced to15 cc/kg 24 hours before shutdown without instituting Action Level 1.Hydrogen concentration never went below 15 cc/kg while critical duringthe reviewed period and is typically maintained between 25 and 50cc/kg. During the reviewed period:
L-2014-252 Attachment Page 13 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0* the dissolved oxygen never exceeded 10 ppb, typically maintained at less than of 5 ppb, and* the concentration of fluoride, chloride and sulfate never exceeded150 ppb, typically maintained at less than 5 ppb.The above values are at power and are for the reviewed period.The reactor coolant system water is analyzed for dissolved oxygen andhalides three times per week with no interval between analysis to exceed72 hours. Analysis for dissolved oxygen is not required when the reactorcoolant system Tavg is less than or equal to 250 degrees F. Analysis forhalides is not required when all fuel is removed from the reactor vesseland the reactor coolant system Tavg is less than 140 degrees F. Thereactor coolant system water is analyzed for sulfate ions at least once per7 days.2. "During the outage in which the half-nozzle or MNSA repairs arescheduled to be implemented, licensees adopting the TR's stresscorrosion crack growth arguments will need to review their plant specificRCS coolant chemistry histories over the last two operating cycles for theirplants and confirm that these conditions have been met over the last twooperating cycles."FPL Assessment:
The contaminant limits, as stated in response toparagraph 1, immediately above, have been maintained at power duringthe review period of June 2005 to March 2014. No transients that exceedthe contaminant concentration limits of paragraph I were identified for thereviewed period.This Relief Request applies to all previous repairs to alloy 600 small bore nozzlesand heater sleeves on the hot leg reactor coolant piping and pressurizer thathave left a remnant nozzle or heater sleeve in place.In conclusion, the ASME Boiler & Pressure Vessel Code Section Xl requirement, IWB-3132.2, is to correct a component containing a flaw. The proposedalternative is to relocate the pressure boundary weld and not correct thecomponent containing the flaw but show by analysis that the material and thepresence of the flaw will not be detrimental to the pressure retaining function ofthe reactor coolant piping and pressurizer.
: Analyses, Reference 1, have shownthat allowing the material containing a flaw to remain in place and in servicewould not result in a reduction of the level of quality or safety.
L-2014-252 Attachment Page 14 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 06. Duration of Proposed Alternative The proposed alternative is for the fourth 10-year Inservice Inspection interval atPSL-2, which began August 8, 2013 and ends August 7, 2023.7. Precedent FPL Relief Request #5, approved by NRC SE Dated May 26, 2006, "St. LucieNuclear Plant, Unit 2 -Regarding Request for Relief from the Requirements ofthe ASME Code (TAC No. MC9502)"
(ML061290056).
: 8. References
: 1) WCAP-15973-P-A, Rev 0 (NRC approved version of WCAP-15973-P, Revision 1with enclosed NRC Safety Evaluation)
"Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Programs",
Westinghouse Electric Company LLC, February 2005 (ML050700431 for NP version)2) NRC letter dated January 12, 2005, "
 
==Subject:==
 
Final Safety Evaluation for TopicalReport WCAP-15973-P, Rev 01 "Low-Alloy Steel Component Corrosion AnalysisSupporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Program"(TAC No. MB6805)"
(ML050180528)
: 3) NRC letter to Mr. J. A. Stall dated August 11, 2005 "St. Lucie Nuclear Plant, Unit1 -Request for Additional Information Regarding Relief Request No. 26 -Repairof Alloy 600 Small Bore Nozzles Without Flaw Removal (TAC No. MC6944)"(ML052210368)
: 4) A-CEOG-9449-1242 Rev. 00 (Task 1131) "Evaluation of the Corrosion Allowance for Reinforcement and Effective Weld to Support Small Alloy 600 NozzleRepairs"5) Westinghouse Calculation Note Number CN-CI-02-69, Rev. 0 "Evaluation ofFatigue Crack Growth Associated with Small Diameter Nozzles for St. Lucie 1 &2" (Non Proprietary Version -ML023380149)
: 6) St. Lucie Unit 2 Updated Final Safety Analysis Report through Amendment No.21 L-2014-252 Attachment Page 15 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0TABLE 1PSL-2 Alloy 600 Small Bore Nozzles Repair StatusLocation Tag ID Repair Repair Method Reason for FlawDate (Figure 1 Design) Repair LeftPZR Stm Space A 1994 1/2 Nozzle Repair* Linear YesUpper Head (B) Indications PZR Stm Space B 1994 1/2 Nozzle Repair* Linear YesUpper Head (B) Indications PZR Stm Space C 1994 1/2 Nozzle Repair* Leakage / Linear YesUpper Head (B) Indications PZR Stm Space D 1994 1/2 Nozzle Repair* Preventative NoUpper Head (B)PZR Wtr Space RC-105 1995 Sleeve Repair* Preventative NoLower Head (C)PZR Wtr Space RC-130 1995 Sleeve Repair* Preventative NoLower Head (C)PZR Wtr Space TE-1 101 1995 Sleeve Repair* Preventative NoSide Shell (C)RCS Hot Leg TE-1112HA 1989 Sleeve Repair* Preventative NoRTD Nozzle (C)RCS Hot Leg TE-1111X 1989 Sleeve Repair* Preventative NoRTD Nozzle (C)RCS Hot Leg TE-1122HC 1989 Sleeve Repair* Preventative NoRTD Nozzle (C)RCS Hot Leg TE-1122HD 1989 Sleeve Repair* Preventative NoRTD Nozzle (C)RCS Hot Leg TE-1121X 1989 Sleeve Repair* Preventative NoRTD Nozzle (C)RCS Hot Leg TE-1112HB 2003 1/2 Nozzle Repair Preventative NoRTD Nozzle (A)RCS Hot Leg TE-1112HC 2003 1/2 Nozzle Repair Preventative NoRTD Nozzle (A)RCS Hot Leg TE-1112HD 2003 1/2 Nozzle Repair Preventative NoRTD Nozzle (A)RCS Hot Leg TE-1 122HA 2003 1/2 Nozzle Repair Preventative NoRTD Nozzle (A)RCS Hot Leg TE-1 122HB 2003 1/2 Nozzle Repair Preventative NoRTD Nozzle I (A) I I L-2014-252 Attachment Page 16 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0TABLE IPSL-2 Alloy 600 Small Bore Nozzles Repair StatusLocation Tag ID Repair Repair Method Reason for FlawDate (Figure 1 Design) Repair LeftRCS Hot Leg PDT-1121B 1995 Sleeve Repair Leakage YesFlow Nozzle (D) LeakageYes RCS Hot Leg PDT-1111A 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg PDT-1111B 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg PDT-11111C 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg PDT-1111D 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg PDT-1121A 1995 Sleeve Repair Preventative NoFlow Nozzle PDT-1121A (0(D) Preventative RCS Hot Leg PDT-1121C 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg PDT-1121D 1995 Sleeve Repair Preventative NoFlow Nozzle (D)RCS Hot Leg Sample Line 1995 Sleeve Repair Preventative NoFlow Nozzle (D)PZR Heater 30 2011 1/2 Nozzle Repair* Preventative NoSleeves I I (B) I I* Nozzle welded to a nickel alloy weld pad.
L-2014-252 Attachment Page 17 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0TABLE 2ASUMMARY OF LIMITING ALLOWABLE DIAMETER CALCULATIONS FOR HALF-NOZZLE REPAIRSNozzle Weld Nozzle Diameter Repair Bore LimitingLocation Joint Repair Bore Corrosion Diameter Allowable Design Diameter Loss After After 49 Diameter(Figure 1) (inch) 49 Years Years (inch)(inch) (inch)Hot Leg Piping A 1.063 0.1313 1.194 1.27Pressurizer B 1.325 0.1313 1.456 2.26Upper HeadPressurizer HeateriSee B 1.693 0.1313 1.824 2.26Heater SleeveTABLE 2BSUMMARY OF LIMITING ALLOWABLE DIAMETER CALCULATIONS FOR SLEEVE REPAIRSNozzle Weld Nozzle Diameter Repair Bore LimitingLocation Joint Repair Bore Corrosion Diameter Allowable Design Diameter Loss After After 54 Diameter(Figure 1) (inch) 54 Years Years (inch)(inch) (inch)C 1.129 0.025 1.154 1.27Hot Leg Piping D 1.178 " 1.203 1.27Pressurizer C 1.5 0.025 1.525 1.62Side Shell C 1.325 and 1.350 2.26Lower Head 1.5 1.525 2.26 L-2014-252 Attachment Page 18 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0TABLE 3SUMMARY OF CHARPY IMPACT DATAName Heat No. Testing *Absorbed
*% *USETemperature Energy Shear ft-lbOF ft-lbReactor Vessel A8490-2 +60 44 23 105PlateReactor Vessel B3416-2 +60 37 20 113PlateReactor Vessel A8490-1 +60 58 28 115PlateReactor Vessel B8307-2 +60 49 22 93PlateReactor Vessel A3131-1 +60 47 22 107PlateReactor Vessel A3131-2 +60 52 23 105PlatePressurizer C4754-3 +70 69 60Bottom HeadPressurizer Upper B8618-2 +70 69 60HeadPressurizer Lower NR 60 466-2 +20 72 35ShellPressurizer Lower NR 61 734-1 +30 54 25Shell* Average of three tests reported in the material certification.
L-2014-252 Attachment Page 19 of 19ST. LUCIE UNIT 2FOURTH INSPECTION INTERVALRELIEF REQUEST NUMBER 2, REV. 0HALF NOZZLE HALF NOZZLE NOZZLE NOZZLEWELD WELDWELD WELDFPAD PADWALL WALL WALL WALLNOZZLE NOZZLEEMNANT EMAT SLEEVE SLEVDESIGN DESIGN DESIGN DESIGN"A" "B" "C" "D"FIGURE 1REPLACEMENT NOZZLE CONFIGURATIONS}}

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