ML062640282
| ML062640282 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/02/2006 |
| From: | Terao D NRC/NRR/ADRO/DORL/LPLIV |
| To: | Rosenblum R Southern California Edison Co |
| Kalyanam N, NRR/DORL, 415-1480 | |
| References | |
| TAC MD9487, TAC MD9488 | |
| Download: ML062640282 (15) | |
Text
October 2, 2006 Mr. Richard M. Rosenblum Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 - RE:
REQUEST FOR RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE (CODE) (TAC NOS. MD9487 AND MD9488)
Dear Mr. Rosenblum:
By letter dated December 23, 2005, Southern California Edison (SCE, the licensee) submitted a request for the use of alternatives to certain ASME Code,Section XI requirements at San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3).
Specifically, Relief Request ISI-3-17 sought relief from the requirements of the 1995 Edition through the 1996 Addenda of ASME Code, Section Xl, IWA-3300, Flaw Characterization, and IWB-2420, Successive Inspections. The alternative described in the proposed relief request will provide an acceptable level of quality and safety when compared to the code requirements in IWB-3500 to characterize the cracks left in service. Therefore, in lieu of fully characterizing any remaining cracks and performing successive examinations to validate flaw stability, SCE utilizes worst-case assumptions to conservatively estimate the crack extent and orientation.
The U.S. Nuclear Regulatory Commission (NRC) staff authorizes the alternative proposed by SCE in accordance with paragraph 50.55a(a)(3)(i) of Title 10 of Code of Federal Regulations, which states that the proposed alternatives may be used when authorized by the Director of the Office of Nuclear Reactor Regulation if the applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety.
R. Rosenblum Therefore, Relief Request ISI-3-17 is authorized for Cycle 14 Refueling Outage for SONGS 2, and mid-cycle outage for SONGS 3. Due to the immediate need of this relief request, verbal authorization for the use of this relief request was granted on March 23, 2006, for SONGS 2, and April 27, 2006, for SONGS 3.
The staff's safety evaluation is enclosed.
Sincerely,
/RA/
David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362
Enclosure:
Safety Evaluation cc w/encl: See next page
- Minor editorial changes made in staff supplied SE OFFICE NRR/LPL4/PM NRR/LPL4/LA DCI/CFEB* OGC - NLO NRR/LPL4/BC NAME NKalyanam LFeizollahi KGruss TCampbell DTerao DATE 9/25/06 9/25/06 8/18/06 9/28/06 10/2/06 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-3-17 SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362
1.0 INTRODUCTION
By letter dated December 23, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML053620021), pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR), Southern California Edison Company (SCE, the licensee) submitted relief request ISI-3-17 to be implemented during the third 10-year inservice inspection (ISI) interval at San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3). In response to verbal requests, the licensee also sent calculations M-DSC-411, M-DSC-414, M-DSC-360, and M-DSC-354 referenced in its application (see References 1, 2, 3, and 4 in Section 5.0 of this safety evaluation for details).
In the relief request, the licensee requested relief from the requirements of paragraph IWA-3300 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, to determine the size of the flaws left in the J-groove weld of the nozzle and instead proposed to use the worst-case assumptions to conservatively estimate the size and orientation of flaw. The licensee also requested relief from IWB-2420 which requires successive inspections of areas of components, containing flaws, that have been accepted for continued service.
2.0 REGULATORY EVALUATION
The ISI requirements of the ASME Code Class 1, Class 2, and Class 3 components in nuclear plants are to be performed in accordance with the ASME Code,Section XI, and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Paragraph 50.55a(a)(3) of 10 CFR states: "Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g),
and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that: (i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
Paragraph 50.55a(b)(1) of 10 CFR provides the requirements that reactor coolant pressure boundary components must meet. This section states that components which are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section III of the ASME Code. This requirement applies to the new repair weld attaching the replacement half-nozzle to the pressurizer shell.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The third 10-year ISI interval for SONGS 2 and 3 began on August 18, 2003, and is scheduled to end on August 17, 2013.
3.0 EVALUATION OF RELIEF REQUEST ISI-3-17 3.1 System/Component(s) for Which Relief is Requested Relief Request ISI-3-17 is related to 3/4-inch lower-level instrument nozzles of the pressurizer.
The nozzles are located in the bottom head of the pressurizer. There are two nozzles per pressurizer in each unit and a total of four for both units. The pressurizer lower-level instrument nozzles and attachment welds are part of the primary pressure boundary and are ASME Code Class 1 components. The nozzles are part of Examination Category B-P, Item No. B15.20, of Table IWB-2500-1 of the ASME Code,Section XI. The examination category applies to the original and modified pressurizer lower-level instrument nozzles.
3.2 Applicable ASME Code ASME Code,Section XI, 1995 Edition through the 1996 Addenda, is the code of record for the third 10-year ISI interval.
3.3 Applicable ASME Code Requirements Subarticle IWA-4611.1 in the 1996 Addenda of ASME Code,Section XI, requires that defects shall be removed or reduced in size. However, Subarticle IWA-4611.1 allows that
...Alternatively, the defect removal area and any remaining portion of the flaw may be evaluated and the component accepted in accordance with the appropriate flaw evaluation rules of Section XI...
ASME Code, Section Xl, Subarticle IWA-3300, requires that flaws detected by inservice examination have their dimensions determined and be characterized.
ASME Code,Section XI, Subarticle IWB-2420 requires existing flaws be reexamined and reevaluated in accordance with Subarticle IWB-3132.3 or IWB-3142.4 of the ASME Code,Section XI, in successive inspection intervals.
3.4 Licensees Alternate Criteria for Acceptability The proposed half-nozzle repair of the pressurizer lower-level instrument nozzles does not include the removal of any flaws assumed to be present in the remnant of the original nozzle or its J-groove partial penetration attachment weld. Subarticle IWA-4611.1 allows the cracks to be evaluated using the appropriate flaw evaluation rules of ASME Code,Section XI. In lieu of characterizing the existing cracks, the licensee assumed a worst-case flaw and evaluated the flaw in accordance with the rules of ASME Code,Section XI, Subarticle IWB-3600.
The licensee requested relief from ASME Code, Section Xl, Subarticle IWA-3300. It is assumed that any flaws present in the original pressurizer lower-level instrument nozzles or J-groove attachment welds will not be removed. In lieu of sizing or characterizing the existing cracks, the licensee assumes worst-case crack extent and orientation.
The licensee requested relief from ASME Code,Section XI, Subarticle IWB-2420, that no additional inspections be performed to monitor the flaw stability because initial flaw sizing and characterization is not being performed so the actual dimensions of the flaw will not be determined. In lieu of crack characterization and successive examinations, the licensee demonstrated the acceptability of the worst-case crack in accordance with the requirements of the ASME Code,Section XI, Subarticle IWB-3600.
3.5 Licensees Basis for Relief The licensee states that removal of the cracks in the existing J-groove partial penetration welds would incur excessive radiation dose for repair personnel. With the installation of the new pressure boundary welds, the function of the original J-groove partial penetration welds is no longer required.
The licensee states that the requirements of Subarticle IWA-4611.1 allow two options for determining the disposition of discovered cracks. The subject cracks are either removed as part of the repair process or left as-is and evaluated per the rules of Subarticle IWB-3600. The proposed repair design dictates that the original nozzle remnant and its attachment weld be left intact inside the vessel. The assumptions of Subarticle IWB-3500 are that the cracks are fully characterized in order to be able to compare the calculated crack parameters to the acceptable parameters provided in IWB-3500. In the proposed alternative, the acceptance of the postulated crack is calculated based on expected crack orientation and the geometry of the weld. Typically, an expected crack orientation is evaluated based on prevalent stresses at the location of interest. Using worst-case (maximum) assumptions with the geometry of the remnant weld, the postulated crack is assumed to begin at the inside surface of the instrument nozzle, penetrate the nozzle wall, continue through to the intersection of the vessel inner-diameter surface and the vessel nozzle bore by primary water stress-corrosion cracking (PWSCC). The crack is assumed to grow into the pressurizer shell wall by fatigue. The depth and orientation are worst-case assumptions for cracks that may occur in the remaining
J-groove partial penetration weld. The licensee assumed that the as left condition of the remaining J-groove weld includes degraded weld and nozzle material.
The licensee performed fracture mechanics evaluations in accordance with ASME Code,Section XI, Subarticles IWB-3600 and IWB-3700. The licensee concluded that degraded J-groove weld material could remain in the vessel, with no examination to size any flaws that might remain following the repair. Since the hoop stresses in the J-groove weld are higher than the axial stresses, the preferential direction for cracking is axial, or radial relative to the nozzle.
It is postulated that a radial crack in the Alloy 182 weld metal would propagate by PWSCC through the weld to the interface with the low-alloy steel shell. It is fully expected that such a crack would then arrest at the weld-to-shell interface.
The licensee states that crack growth through the Alloy 182 material would tend to relieve the residual stresses in the weld as the crack grows to its final size. Although residual stresses in the shell material are low, it is assumed that a weld flaw formed by PWSCC could continue to extend into the low-alloy steel material due to cyclic loading. This flaw will form a continuous radial-comer flaw that would propagate into the low-alloy steel shell by fatigue-crack growth under cyclic loading conditions.
Hoop stresses are used since they are perpendicular to the plane of the crack. The life of the repair is determined based on fatigue-crack growth of 40 years of additional service. The final flaw size meets the fracture toughness requirements of the ASME Code using an upper-shelf value of 200 ksi /in for unirradiated ferritic materials. The results of the analyses indicate that it is acceptable to leave the original Alloy 600 instrument nozzle remnant and original attachment J-groove weld in the pressurizer, even with the possibility that cracks exist in the weld for 40 years of service.
The licensee also considered the possibility of a transverse crack and a radial crack intersecting and generating a loose part in the pressurizer. As noted above, radial cracks are postulated to occur in the weld due to the dominance of the hoop stress at this location. The licensee states that only thermal and welding residual stresses could cause a transverse crack to grow.
However, the presence of radial cracks limits the growth potential of the transverse cracks. The radial cracks would relieve the potential transverse crack driving forces. Hence, it is unlikely that a series of transverse cracks could intersect a series of radial cracks resulting in any fragments becoming dislodged. Therefore, the release of debris generated by a cracked weld is highly unlikely.
The licensee evaluated the consequences of loose parts of similar or larger dimensions and mass being carried into the reactor vessel and concluded that the probability of damage to any reactor coolant system component is not significant. In the unlikely event that a small part of the heater sleeve remnant breaks away, the density of the loose part material is significantly greater than that of the water, and the loose part would tend to settle to the bottom of the pressurizer.
Flow velocities in the pressurizer are relatively low and are not likely to transport loose pieces of metal to the surge line. Additionally, the surge-line nozzle is equipped with a surge screen, with half-inch holes. The licensee concluded that no damage is expected in the event that the loose part does not remain in the bottom of the reactor.
3.6 Implementation Schedule The licensee proposed to implement Relief Request ISI-3-17 during the third ISI interval, which began on August 18, 2003, and is scheduled to end on August 17, 2013.
3.7 Staffs Evaluation The licensees proposed alternative for the repair of the pressurizer lower-level instrument nozzles is based on several flaw evaluations and stress analyses that were submitted with Relief Request ISI-3-17. In lieu of requirements of Subarticles IWA-3300 and IWB-2420, the licensee performed a flaw evaluation in accordance with Subarticle IWB-3600 and a supplemental calculation based on elastic-plastic fracture mechanics. The staff reviewed the licensees flaw evaluations and stress analyses to verify that the proposed alternative would provide an acceptable level of quality and safety for the operation of the pressurizer.
The licensee calculated the effect of residual stresses due to welding and thermal transient stresses in the existing J-groove weld of the pressurizer bottom head as discussed in Reference 1. The licensee simulated the residual stress associated with fabricating the lower-level nozzle penetration repair. The simulation included loading due to welding, thermal stress relief, hydrostatic testing, operating conditions, final residual stress, and the repair. The results show that the high tensile hoop stresses in the weld and buttering dissipates and becomes compressive within a short distance into the pressurizer base metal from the interface between the butter and pressurizer shell. The compressive stresses are favorable in this situation because they would restrict the propagation of the postulated flaw in the pressurizer shell. The results of the residual stress and thermal transient analyses were used to evaluate flaw stability in the fracture mechanics calculations discussed below.
Using the residual and thermal stresses calculated from Reference 1, the licensee analyzed the impact of an assumed flaw in the existing J-groove weld propagating into low-alloy steel base material as discussed in Reference 2. Attachment I to Reference 2 is related to the flaw evaluation of the pressurizer lower-level nozzle repair by a licensees contractor, Dominion Engineering. Attachment II to Reference 2 is related to the flaw evaluation of the pressurizer thermowell nozzle repair by Dominion Engineering. Attachment III to Reference 2 is an independent flaw evaluation of the lower-level and thermowell nozzle repairs by another licensees contractor, Aptech Engineering. The staff reviewed Attachment I to Reference 2 only. The staff did not review Attachments II and III to Reference 2 because Attachment II is related to thermowell nozzles which is not part of Relief Request ISI-3-17. The staff also did not review Attachment III because Attachment III is an independent verification of calculations presented in Attachments I and II.
In Attachment I to Reference 2, the licensee assumed an initial flaw starting at the inside surface of the remnant nozzle. The initial flaw is assumed to propagate through the nozzle wall into the J-groove weld and butter due to PWSCC, (i.e., the entire J-groove weld was assumed cracked). The crack tip is assumed to be located at the interface between the butter and the pressurizer shell base metal. The licensee assumed that the flaw would propagate into the pressurizer base metal by fatigue. The licensee evaluated the postulated flaw at the uphill and downhill locations on a plane perpendicular to the circumferential direction of the pressurizer
lower head. The flaw was calculated to grow in the pressurizer shell for 40 years following the weld overlay installation using design pressure and temperature transient loading.
Once the final flaw crack size was determined, the licensee used elastic-plastic fracture mechanics to determine the flaw stability in the pressurizer shell. The flaw stability is based on the J-integral approach and safety factors of 3 on the pressure and 1.5 on thermal and residual loads. The licensee's calculation showed that the applied J-value for the final flaw size is less than the J-value of the base metal by a factor of more than 3.00. This means that the postulated flaw will be stable and that the pressurizer base metal has sufficient fracture toughness to resist rapid propagation of the postulated flaw. The licensee concluded that a postulated crack growth through the entire, existing J-groove weld would be within the allowable limits, which is acceptable for 40 years following the weld overlay installation.
For the licensees residual stress analysis, the staff determined that the finite element model of the pressurizer head nozzle geometry is consistent with the as-built drawing of the nozzle. The staff reviewed various transients that were used in the licensees analysis, including heatup, cooldown with flooding, reactor trip, loss of secondary pressure, loading/unloading, and leak test. The staff determined that the transients are consistent with the design basis. The staff also verified that the material specifications and associated material properties used in the licensees analyses are consistent with the design specification. The licensee stated that the finite element model was verified by comparing the analytical results of the model with the experimental and field data that were available at the time. The verification study of the finite element model showed that the locations of observed cracking correlated well with regions of highest stress in the analytical model. The licensees verification study shows that the finite element model is acceptable to be used in the SONGS application. The staff finds that the licensees residual stress evaluation is acceptable because the appropriate modeling and loading conditions were included in the analysis.
The licensee performed linear elastic-fracture mechanics calculations in accordance with Subarticle IWB-3600 of the ASME Code,Section XI. The licensee also used elastic-plastic fracture mechanics to demonstrate the stability of the postulated flaw. The staff evaluated the licensees flaw evaluation and focused on the following four issues.
The first issue is related to the licensees use of elastic-plastic fracture mechanics in the flaw-stability calculation. The staff noted that licensees who have repaired nozzle cracking of their reactor vessels and pressurizers have found that the postulated flaw in the reactor vessels or pressurizers could not satisfy certain safety margins of Subarticle IWB-3600. As an alternative, licensees have used elastic-plastic fracture mechanics to demonstrate flaw stability.
The staff has approved the use of elastic-plastic fracture mechanics in lieu of linear elastic-fracture mechanics (as required in the ASME Code, Subarticle IWB-3600) to demonstrate flaw stability in the reactor vessel and pressurizer (References 5 and 6, respectively).
In the staffs review of flaw-stability calculations in the reactor vessel of Arkansas Nuclear One, Unit 1 (ANO-1) nuclear plant (Reference 5) and the pressurizer of Palo Verde nuclear plant (Reference 6), the staff found that the licensees calculated safety factors of the postulated crack in the J-groove weld were less than the required safety factor of /10 as specified in Subarticle IWB-3613 of the ASME Code,Section XI. The low calculated safety factors were a
result of a conservatism which stems from the magnitude and modeling of the weld residual stresses as a constant applied load on the crack face. Another conservatism stems from the use of linear elastic-fracture mechanics. The linear elastic-fracture mechanics of Subarticle IWB-3600 is inherently conservative as applied to low-alloy steel of the reactor vessel and pressurizer because linear elastic-fracture mechanics does not consider the ductility of low-alloy steel. The licensees low calculated safety factors do not imply that the structural integrity of the reactor vessel is compromised but rather, is a reflection of the conservatism in the analytical approach and assumptions. Therefore, the staff allowed the use of elastic-plastic fracture mechanics to demonstrate the flaw-stability in the reactor vessel and pressurizer because the staff concluded that elastic-plastic fracture mechanics is appropriate as applied to low-alloy steel of the reactor vessel and pressurizer for the specific flaw configuration and loading. The licensee for SONGS has shown that the calculated safety factor for the postulated flaw in the SONGS pressurizer shell also was less than that required by ASME Code safety margins of IWB-3600. The staff finds that the flaw configuration and assumptions for the SONGS half-nozzle repair is similar to the flaw configuration and assumptions in References 5 and 6.
Therefore, it is acceptable for SONGS to use elastic-plastic fracture mechanics to demonstrate the flaw stability in the SONGS pressurizer shell.
The second issue is related to the licensees use of a safety factor of /2 for the cooldown transient event in its flaw evaluation, to satisfy the requirements of Subarticle IWB-3600. The licensee stated that use of a safety factor of /2 for the normal and upset conditions, when the internal pressure is less than 20 percent, was suggested by ASME Code interpretation XI-1-04-03. The staff does not recognize ASME Code interpretations and does not agree with the licensees use of a safety factor of /2 on the stress intensity factor for the cooldown transient event. As required in ASME Code, Subarticle IWB-3612, a safety factor of /10 must be applied to the stress intensity factor derived under the normal and upset load conditions.
However, IWB-3613(a) allows the following:
The following criteria shall be used for the evaluation of flaws in areas of structural discontinuity, such as vessel-flange and nozzle-to-shell regions. A flaw exceeding the limits of IWB-3500 is acceptable if the applied stress intensity factor for the dimensions af and lf satisfies the following limits.
(a) For conditions where pressurization does not exceed 20% of Design Pressure, during which the minimum temperature is not less thean RTNDT:
KI<KICI%2 (The variables are defined in IWB-3613(a))
In addition, the licensee used the KIC value per Subarticle IWB-3613 of the 2005 Addenda to the 2004 Edition of the ASME Code,Section XI, for the allowable crack tip stress intensity factor at zero-load conditions. The value of KIa is required in Subarticle IWB-3613 of the 1995 Edition of the ASME Code and is more conservative than the KIC value. The staff has not approved the use of the 2005 Addenda to the 2004 Edition of ASME Code,Section XI. Also, licensees who wish to use the editions of the ASME Code later that the edition that is approved for the ISI interval need to request the NRC's approval prior to the use as specified in NRC
Regulatory Issue Summary 2004-16. Nevertheless, the licensee has demonstrated flaw stability in the pressurizer shell based on elastic-plastic fracture mechanics. The licensee used the more precise elastic-plastic mechanics technique to demonstrate flaw stability which is acceptable for this application. Hence the use of safety factor of /2 per ASME Code interpretation XI-1-04-03 and use of the KIC value instead of KIa become not relevant.
The third issue is related to whether the base metal of the SONGS 2 pressurizer bottom shell has sufficient resistance to crack propagation. Resistance to crack propagation depends on the fracture toughness of a material which, in this case, is measured by the upper-shelf Charpy energy of the pressurizer base metal. Because of the vintage of the SONGS plant, the licensee has no upper-shelf Charpy energy data for the pressurizer base metal, which is usually obtained at high-test temperature (e.g., >150 "F) in the Charpy V-notch tests. The available Charpy energy data for the pressurizer shell were taken at a relatively low temperature (+10 oF) and in the plate longitudinal (rolling) direction. However, in the flaw-stability calculations, the Charpy energy values at the transverse direction are needed. The licensee converted the Charpy energy in the longitudinal direction to the transverse direction in accordance with the procedures described in Standard Review Plan Section 5.3.2, Pressure-Temperature Limits.
The equivalent Charpy energy values for SONGS 2 pressurizer shell material at +10 oF are 60, 64, and 66 ft-lbs (at 70 percent shear fracture appearance for all tests). For SONGS 3, the equivalent Charpy energy values are 22, 23, and 25 ft-lbs (at 30 percent shear fracture appearance for all tests).
As discussed in Reference 5, to support the ANO-1 submittal, Entergy provided the complete Charpy energy versus temperature data from the Grand Gulf reactor vessel material. Entergy stated that the materials of the ANO-1 and Grand Gulf reactor vessel head are comparable.
The plate materials for both plants are SA-533, Grade B, Class 1 and were supplied by the same steel company, Lukens Steel. Entergy superimposed the ANO-1 Charpy energy data on the Grand Gulf Charpy energy graph and showed that the ANO-1 data fit the general trend of the Grand Gulf Charpy energy data. For the upper-shelf Charpy energy for ANO-1, Entergy extrapolated a lower-bound Charpy energy of 94 ft-lbs from the Grand Gulf data.
As discussed in Reference 5, in the evaluation of the ANO-1 submittal, the staff performed an independent analysis of the Charpy energy values by reviewing surveillance capsule reports of reactor vessel material specimens and final safety analysis reports of various nuclear plants.
The staff found that the ANO-1 Charpy test data are consistent with the Charpy energy data from other Babcock and Wilcox reactor vessels with the same SA-533 plate material. The staff also found in Electric Power Research Institute report, TR-113596, that the upper-shelf Charpy energies of SA-533 plate materials fabricated by Lukens Steel between 1966 and 1974 ranged from 91 ft-lbs to 177 ft-lbs. The mean upper-shelf energy was calculated to be 127 ft-lbs with one standard deviation of 15 ft-lbs. The upper-shelf energy at 2 standard deviations from the mean was 97 ft-lbs, which represents the 95-percent confidence interval value. The lower the upper-shelf energy used in the elastic-plastic fracture mechanics calculation, the more conservative the flaw-stability results will be. That is, if a crack can be shown to be stabilized at a lower upper-shelf energy value, the crack will become stabilized at a higher upper-shelf energy value. Therefore, the staff found that an upper-shelf energy of 94 ft-lbs used by Entergy was acceptable because the 94 ft-lbs energy value is conservative as compared to the 95-percent confidence value of 97 ft-lbs.
The licensee stated that the SONGS pressurizer was fabricated by the same Lukens Steel company and pressurizer base metal is made of the same SA-533, Grade B, Class 1 low-alloy steel as that of the ANO-1 reactor vessel. On the basis of the staffs acceptance of the 94 ft-lbs upper-shelf energy value for ANO-1's reactor vessel, the licensee used 94 ft-lbs for the SONGS pressurizer. The staff finds that it is acceptable for SONGS to use an upper-shelf energy of 94 ft-lbs in its flaw-stability calculation.
The fourth issue is related to the licensees proposed safety factors (i.e., safety factors of 3 on the pressure and 1.5 on thermal loading) for its elastic-plastic fracture mechanics calculations.
The subject safety factors are lower than the ASME Code,Section XI, safety factor of /10 (as specified in IWB-3613) because the same safety factor, /10, is applied to the stress intensity factor derived from the primary stresses (pressure) and secondary stresses (thermal).
The licensee believes that this results in an overly-conservative allowable stress-intensity factor when the predominant loading mechanism in the J-groove weld is highly localized and due to residual stresses, which are considered as the secondary stresses. The NRC staff reviewed the use of the proposed safety factor of 3 on pressure loading and 1.5 on thermal loading in the approval of ANO-1's nozzle repair and found it acceptable. Since the flaw configuration and loading are similar between SONGS 2 and 3 and ANO-1, the NRC staff finds that the proposed safety factors are acceptable for SONGS 2 and 3 and concludes that the proposed safety factor will provide adequate margin for the structural integrity of the pressurizer bottom head.
The staff finds that the licensees flaw-stability evaluation is acceptable because the evaluation assumed a worst-case flaw originating in the nozzle, propagating through the J-groove weld and into the pressurizer base metal. The staff concludes that the licensee has demonstrated by analysis that the proposed half-nozzle repair will not affect the structural integrity of the pressurizer base metal significantly.
In addition to the flaw-stability evaluation discussed above, the licensee also performed an analysis to demonstrate the long-term acceptability of the half-nozzle configuration, considering the potential corrosion of the low-alloy steel shell material which will be exposed to primary system borated water (Reference 3). The half-nozzle repair will result in a 1/16-inch gap inside the pressurizer nozzle (bore) penetration between the original nozzle stub and the new half nozzle. The primary coolant could become trapped in the gap which could cause corrosion of the pressurizer base metal inside the nozzle penetration. To evaluate the potential for corrosion, the licensee modeled local corrosion as a circumferential planar groove within the pressurizer nozzle penetration. The groove represents the 1/16-inch gap between the original nozzle stub and the new half nozzle inside the pressurizer nozzle penetration. The degradation would consist of the loss of metal in the annulus between the nozzle and shell penetration. The integrity of the nozzle attachment was determined as a function of the location of borated water corrosion within the pressurizer bore penetration, and depth and length of the gap. The licensee also simulated a crevice between the weld attachment pad and the pressurizer shell in which localized boric water corrosion was assumed to occur. The licensee used the general approach of ASME Code,Section XI, Appendix H, to perform a limit load-based evaluation including fatigue-crack growth on the nozzle in the pressurizer. The allowable corrosion depths were computed at the gap region between the new nozzle and the remnant nozzle stub, and in the crevice region at the nozzle-to-pad weld. The licensee demonstrated that the projected corrosion depths in the gap and crevice regions are within the allowable corrosion depths after 40 years. The NRC staff accepts the approach taken and the conclusions reached by the
licensee to demonstrate that the potential for borated water corrosion as a result of the half-nozzle repair will not affect the integrity of the pressurizer bottom head.
The licensee requested relief from Subarticle IWA-4611.1 of the 1995 Edition through the 1996 Addenda of ASME Code,Section XI, which requires that defects be removed or reduced in size. However, Subarticle IWA-4611.1 allows a flaw evaluation to be performed in lieu of defect removal. The licensee has performed flaw evaluations to satisfy Subarticle IWA-4611.1.
Subarticle IWA-3300 of ASME Code,Section XI, requires characterization of the size of the flaw if it was found during an inservice examination. Subarticle IWB-2420 requires existing flaws be reexamined and reevaluated in accordance with Subarticle IWB-3132.2 or IWB-3142.4 of the 1995 Edition through the 1996 Addenda of ASME Code,Section XI. Since it is difficult to examine the remnant J-groove welds and remnant nozzle stubs which are located inside the pressurizer penetration, the licensee proposed to use an analysis to demonstrate the flaw stability in lieu of inservice examinations. The licensee assumed a worst-case flaw and did a flaw-evaluation and demonstrated that the structural integrity of the pressurizer shell will be maintained.
Since the licensee has been able to demonstrate that the structural integrity of the pressurizer shell will be maintained with the worst-case analytical evaluation technique, the NRC staff accepts that subsequent inservice examination of the J-groove attachment weld as required by Subarticles IWA-3300 and IWB-2420 would not be needed.
The NRC staff concludes that the structural integrity of the pressurizer is demonstrated to be adequate and that the borated water corrosion will not affect the pressurizer shell base metal significantly.
4.0 CONCLUSION
S On the basis of the above evaluation, the NRC staff concludes that the proposed alternative in Relief Request ISI-3-17, pertaining to the remnant J-groove welds of the low-level instrument nozzles in the pressurizer at SONGS 2 and 3, will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the staff authorizes the proposed Relief Request ISI-3-17 for the remaining third 10-year ISI interval at SONGS 2 and 3.
All other requirements of ASME Code,Section III and XI, for which relief has not been specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
5.0 REFERENCES
- 1. M-DSC-411, Rev. 0, SONGS Unit 2 AND 3 Pressurize Lower Level Nozzle Welding and Transient Analysis (ADAMS Accession No. ML061090482).
- 2. M-DSC-414, Rev. 0, SONGS Unit 2 & 3 Pressurize Lower Level and Thermal Nozzles J-Weld Evaluation (ADAMS Accession No. ML061090522).
- 3. M-DSC-360, Rev. 0, Evaluation of Half-Nozzle Repair for Pressurize and Steam Generator Inst. Nozzles Under Long-Term Service Conditions SONGS 2 and 3 (ADAMS Accession No. ML061090479).
- 4. M-DSC-354, Rev. 1, Pressurizer Bottom Head Instrumentation Level Nozzle Evaluation (ADAMS Accession No. ML062280369).
- 5. NRC Letter to Entergy dated September 29, 2004, Safety Evaluation by the Office of Nuclear Reactor Regulation Inservice Inspection Program - Relief Requests Nos. ANO1-R&R-005 and ANO1-R&R-006 (ADAMS Accession No. ML042890174).
- 6. NRC letter to Arizona Public Service Company dated November 5, 2004, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Relief Request No. 29 RE: Remnant Sleeves(s) Flaw Evaluation (TAC Nos. MC3606, MC3607, and MC3608) (ADAMS Accession No. ML043130170).
Principal Contributors: Edward Andruszkiewicz John Tsao Date: October 2, 2006
San Onofre Nuclear Generating Station Units 2 and 3 cc:
Mr. Daniel P. Breig Resident Inspector/San Onofre NPS Southern California Edison Company c/o U.S. Nuclear Regulatory Commission San Onofre Nuclear Generating Station Post Office Box 4329 P. O. Box 128 San Clemente, CA 92674 San Clemente, CA 92674-0128 Mayor Mr. Douglas K. Porter, Esquire City of San Clemente Southern California Edison Company 100 Avenida Presidio 2244 Walnut Grove Avenue San Clemente, CA 92672 Rosemead, CA 91770 Mr. James T. Reilly Mr. David Spath, Chief Southern California Edison Company Division of Drinking Water and San Onofre Nuclear Generating Station Environmental Management P.O. Box 128 P. O. Box 942732 San Clemente, CA 92674-0128 Sacramento, CA 94234-7320 Mr. James D. Boyd, Commissioner Chairman, Board of Supervisors California Energy Commission County of San Diego 1516 Ninth Street (MS 31) 1600 Pacific Highway, Room 335 Sacramento, CA 95814 San Diego, CA 92101 Mr. Ray Waldo, Vice President Mark L. Parsons Southern California Edison Company Deputy City Attorney San Onofre Nuclear Generating Station City of Riverside P.O. Box 128 3900 Main Street San Clemente, CA 92764-0128 Riverside, CA 92522 Mr. Brian Katz Mr. Gary L. Nolff Southern California Edison Company Assistant Director - Resources San Onofre Nuclear Generating Station City of Riverside P.O. Box 128 3900 Main Street San Clemente, CA 92764-0128 Riverside, CA 92522 Mr. Steve Hsu Regional Administrator, Region IV Department of Health Services U.S. Nuclear Regulatory Commission Radiologic Health Branch 611 Ryan Plaza Drive, Suite 400 MS 7610, P.O. Box 997414 Arlington, TX 76011-8064 Sacramento, CA 95899 Mr. Michael R. Olson Mr. A. Edward Scherer San Diego Gas & Electric Company Southern California Edison Company 8315 Century Park Ct. CP21G San Onofre Nuclear Generating Station San Diego, CA 92123-1548 P.O. Box 128 San Clemente, CA 92674-0128 Director, Radiologic Health Branch State Department of Health Services P.O. Box 997414, MS 7610 Sacramento, CA 95899-7414 March 2006