ML080250491

Request for Alternative W3-ISI-003 to ASME Code Requirements to Allow Extension to Second 10-Year Inservice Inspection Interval Beyond That Currently Allowed by ASME Subsection IWA-2340(d)

ML080250491
Person / Time
Site:	Waterford
Issue date:	02/15/2008
From:	Hiltz T NRC/NRR/ADRO/DORL/LPLIV
To:	Walsh K Entergy Nuclear Operations
Kalyanam N, NRR/DORL/LPL4, 415-1480
References
TAC MD5387
Download: ML080250491 (12)

Text

February 15, 2008 Mr. Kevin T. Walsh Vice President, Operations Entergy Nuclear Operations 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - REQUEST FOR ALTERNATIVE W3-ISI-003 FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI (TAC NO. MD5387)

Dear Mr. Walsh:

By letter dated April 26, 2007, Entergy Operations, Inc. (the licensee), submitted, pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR), Request for Alternative W3-ISI-003, proposing an alternative to the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

The licensees request is to allow for an extension to the Waterford Steam Electric Station, Unit 3 (Waterford 3) second 10-year inservice inspection (ISI) interval beyond that currently allowed by the ASME Code under Subsection IWA-2430(d). Specifically, Entergy proposes to perform the second 10-year ISI reactor vessel (RV) weld examination during Waterford 3s fall 2009 refueling outage.

The U.S. Nuclear Regulatory Commission (NRC) staff finds that operation of the RV for an additional cycle, without performing the ISI examination of the subject welds, would not significantly increase the risk of either flaw growth due to fatigue or RV failure due to pressurized thermal shock events. Therefore, the staff finds that the licensee=s proposed alternative provides an acceptable level of quality and safety and, therefore, the alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i). The staff authorizes the proposed alternative for the second 10-year ISI interval at Waterford 3, until the end of the fall 2009 refueling outage (RF16) for Waterford 3.

All other requirements of the ASME Code for which relief has not been specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

K. T. Walsh The NRC staff's safety evaluation is enclosed.

Sincerely,

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosure:

Safety Evaluation cc w/encl: See next page

ML080250491 NRR-028 *Minor editorial changes made from staff provided SE

• • NLO w/edits OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/DCI/CVIB* OGC** NRR/LPL4/BC NAME NKalyanam:sp JBurkhardt MMitchell* PMoulding** THiltz DATE 2/4/08 2/4/08 1/22/08 2/13/08 2/15/08 Waterford Steam Electric Station, Unit 3 (2/12/2008) cc:

Senior Vice President Mr. Timothy Pflieger Entergy Nuclear Operations Environmental Scientist - Supervisor P.O. Box 31995 REP&R-CAP-SPOC Jackson, MS 39286-1995 Louisiana Department of Environmental Quality Vice President, Oversight P.O. Box 4312 Entergy Nuclear Operations Baton Rouge, LA 70821-4312 P.O. Box 31995 Jackson, MS 39286-1995 Parish President Council St. Charles Parish Senior Manager, Nuclear Safety P.O. Box 302

& Licensing Hahnville, LA 70057 Entergy Nuclear Operations P.O. Box 31995 Chairman Jackson, MS 39286-1995 Louisiana Public Services Commission P.O. Box 91154 Senior Vice President Baton Rouge, LA 70825-1697

& Chief Operating Officer Entergy Operations, Inc. Mr. Richard Penrod, Senior Environmental P.O. Box 31995 Scientist/State Liaison Officer Jackson, MS 39286-1995 Office of Environmental Services Northwestern State University Assistant General Counsel Russell Hall, Room 201 Entergy Nuclear Operations Natchitoches, LA 71497 P.O. Box 31995 Jackson, MS 39286-1995 Resident Inspector Waterford NPS Manager, Licensing P.O. Box 822 Entergy Nuclear Operations Killona, LA 70057-0751 Waterford Steam Electric Station, Unit 3 17265 River Road Regional Administrator, Region IV Killona, LA 70057-3093 U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. W3-ISI-003 ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By letter dated April 26, 2007, Entergy Operations, Inc. (Entergy, the licensee) submitted Request for Alternative W3-ISI-003 (Reference 1) for Waterford Steam Electric Station, Unit 3 (Waterford 3), wherein the licensee requested an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI to allow for an extension to the Waterford 3 second 10-year inservice inspection (ISI) interval with respect to certain reactor vessel weld examinations.

2.0 REGULATORY REQUIREMENTS ISI of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(g), except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i).

Paragraph 50.55a(a)(3) of 10 CFR states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service 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 regulation requires 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 applicable Code of record for the second 10-year ISI program interval at Waterford 3 is the 1992 Edition of the ASME Code,Section XI, without addenda.

ENCLOSURE

The second 10-year ISI program interval at Waterford 3 began on July 1, 1997, and was originally scheduled to end on June 30, 2007. Subsection IWA-2430(a) of the 1992 Edition of the ASME Code states: "[t]he inservice examinations and system pressure tests required by IWB, IWC, IWD, and IWE shall be completed during each of the inspection intervals for the service lifetime of the power unit. The inspections shall be performed in accordance with the schedules of Inspection Program A of IWA-2431, or, optionally, Inspection Program B of IWA-2432." Subsection IWB-2410 of this edition of the ASME Code states: "[i]nservice examination and system pressure tests may be performed during the plant outages such as refueling shutdowns or maintenance shutdowns. Waterford 3 has adopted Inspection Program B of IWA-2432 per the inspection scheduling requirements of IWB-2412.

Subsection IWA-2430(d) of the relevant ASME Code states, "[f]or components inspected under Program B, each of the inspection intervals may be extended or decreased by as much as 1 year. Adjustments shall not cause successive intervals to be altered by more than 1 year from the original pattern of the intervals." In accordance with this ASME Code-allowed extension, Entergy has opted to use this provision, thus extending the end of the second 10-year ISI interval to June 30, 2008. However, Entergy proposes to perform the second 10-year ISI interval reactor vessel (RV) weld examinations during the unit=s fall 2009 refueling outage, which is approximately 17 months beyond the ASME Code-allowed 1-year extension. Therefore, Entergy has submitted Request for Alternative W3-ISI-003, which proposes an additional extension of the second 10-year ISI interval to the end of the fall 2009 refueling outage for the subject RV welds.

3.0 EVALUATION 3.1 Component Identification Request for Alternative W3-ISI-003 addresses the following ASME Code,Section XI, Examination Categories and Item Numbers covering examinations of Class 1 components.

These Examination Categories and Item Numbers are from Table IWB-2500-1 of the 1992 Edition of ASME Code,Section XI as follows:

Examination Category Item Number Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Shell Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-A B1.50 Repair Welds B-A B1.51 Beltline Region Repair Welds B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas Circumferential Welds in Piping (only for the reactor vessel inlet and B-J B9.11 outlet nozzle to piping welds)

3.2 Licensee=s Basis for Proposed Alternative The proposed technical basis for extending the second 10-year RV ISI interval to the end of the fall 2009 refueling outage is contained in a letter from R. Gramm of the NRC to G. Bischoff of the Westinghouse Owners Group, dated January 27, 2005 (Reference 2). This letter identifies the five areas that form the basis for the technical justification. Entergy has addressed all five technical areas in its April 26, 2007, submittal as follows:

1. Plant-specific RV ISI history

2. Pressurized Water Reactor (PWR) RV ISI history

3. Degradation mechanisms in the RV

4. Material condition of the RV relative to embrittlement

5. Operational experience relative to RV structural integrity challenging events Waterford 3 is currently in its second 10-year ISI interval for the RV weld examinations. The preservice examination and one ISI examination have been performed on the Examination Category B-A, B-D, and B-J welds to date for Waterford 3.

The preservice and first 10-year ISI interval examinations were performed in accordance with ASME Code,Section XI, and Regulatory Guide (RG) 1.150. These examinations achieved essentially 100 percent coverage for the majority of the welds. No reportable indications were found for any of these welds. The licensee provided tables in its submittal showing a detailed inspection history for the welds and claims that, due to the examination method used and the coverage obtained on the welds, any significant flaws that could challenge RV integrity would have been detected by the preservice and first 10-year ISI interval examinations.

Pursuant to the technical basis criteria described in Reference 2, the licensee conducted a survey of the RV ISI history for 14 PWRs representing 301 total years of service and included RVs fabricated by various vendors. None of the 14 plants surveyed reported any findings during examinations of Category B-A, B-D, and B-J welds. All PWR plants except one have performed their first 10-year ISI of the subject welds and no surface-breaking or unacceptable near-surface flaws have been reported in any of these inspections, which were performed pursuant to the provisions of RG 1.150 or the ASME Code,Section XI, Appendix VIII.

The licensee has also conducted an assessment of the possible RV degradation mechanisms.

According to the licensee, the only currently known degradation mechanism for the subject welds is fatigue due to thermal and mechanical cycling from operational transients. Based on flaw growth simulation studies, Entergy identified the cooldown transient as having the greatest contribution to flaw growth. Based on the low likelihood of more than one or two cooldown transients occurring during the proposed ISI interval extension period and the relatively low fatigue usage factors for the subject welds, the licensee concluded that any flaw growth due to fatigue during the proposed ISI interval extension period is expected to be inherently small.

Entergy noted that from a loading perspective, the most severe operational challenge to RV integrity is due to pressurized thermal shock (PTS) events. The licensee stated that the Waterford 3 RV weld materials are below, and will remain below, the PTS screening criteria (according to 10 CFR 50.61) during the proposed second ISI interval extension period. The licensee provided a table depicting the projected PTS reference temperature (RTPTS) values for each of the RV beltline materials. According to the licensee, this table demonstrates that the RTPTS values for all RV beltline materials will remain below the PTS screening criteria of 10 CFR 50.61.

Entergy has stated that Waterford 3 has implemented emergency operating procedures (EOPs) and operator training to provide assurance that the likelihood of a severe PTS event over the next operating cycle is very low. The operator training stresses fundamental EOP coping strategies in both the classroom and simulator forums. Included in the curriculum are procedure entry conditions, floating steps, fundamental rules, mitigation strategies, time-critical actions, and background information from the basis documents. Critical tasks chosen to be of the utmost importance typically include the preservation and protection of fission product barriers.

Entergy characterized the plant=s response to three scenarios (developed by the NRC staff during its PTS risk reevaluation work) believed to be the most likely scenarios that could cause a PTS event that would challenge significant flaws in the RV welds. The three scenarios are initiated by the following infrequent events:

1. Any transient with reactor trip followed by one stuck-open pressurizer safety/relief valve (PSRV) that recloses after about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Severe PTS events also require the failure to properly control high-pressure injection (HPI).

Entergy stated that when reactor coolant system (RCS) pressure continues to decrease following a trip because of a PSRV opening and sticking open, engineered safety features actuation system (ESFAS) actuation would occur and the operators would enter the ESAS procedure. While the PSRV remains open, the ESFAS procedure provides guidance to control RCS low pressure within the limits of the RCS pressure-temperature (PT) limit curves, provided sub-cooled margin is adequate. Assuming the PSRV eventually closes, operators would recognize rising RCS pressure, validate that HPI throttling criteria had been met, and throttle or secure HPI to prevent system repressurization.

2. Large loss of secondary steam from steamline break or stuck-open atmospheric dump valves. Severe PTS events also require the failure to properly control auxiliary feedwater flow rate and destination (e.g., away from the affected steam generator) and failure to properly control HPI.

Entergy stated that decreasing steam generator pressure following a trip would result in main steamline isolation (MSLI) and emergency feedwater (EFW) actuation. These systems would function to isolate the affected steam generator and provide EFW to the unaffected steam generator. Entergy identified the procedure that provides guidance to the operators for verifying proper MSLI and EFW response including contingency actions to manually control the systems in the event of system or component malfunction. Energy stated that operators would transition to an overcooling procedure which directs initiation of HPI to

make up for RCS inventory shrinkage. The procedure directs the operator to stabilize RCS temperature using the cooling available by steaming the unaffected steam generator and provides direction to secure HPI, provided RCS inventory is satisfactory.

3. A small-break loss-of-coolant accident (for piping with nominal diameters ranging from 4 to 9 inches) that exceeds normal makeup capacity. Severity of a PTS event depends on the break location and the primary injection system's flow rate and water temperature.

Entergy stated that the EOPs and operator training at Waterford 3 are symptom-based and provide the necessary guidance to supplement cooling by use of primary-to-secondary heat transfer with the steam generators. The EOPs provide guidance to bound all break spectrums in order to maintain RCS pressure and temperature within the limits of the applicable PT limit curves.

3.3 Staff Evaluation Entergy summarized prior ISI examinations performed on the RV welds for Waterford 3. All of the subject welds have been examined with no indications reported. The fundamental purpose of performing RV weld ISI examinations is to detect and size flaws, thereby predicting subsequent flaw growth before the flaws grow to the critical dimensions prerequisite of failure.

Although ultrasonic examination technology has improved over the past decades, the geometry and materials involved in RV weld examinations are such that these exams have not been particularly challenging from an inspection-technology perspective. Therefore, the staff agrees with Entergy=s qualitative assessment that the prior examinations were of sufficient quality to identify any significant flaws that would challenge RV integrity and that no significant flaws have been identified.

The licensee discussed the population of all PWRs and indicated that no surface-breaking flaws have been discovered and, for a population of 14 plants that were reviewed in detail, no reportable indications were identified in any of the RV welds. Therefore, the staff concludes that the fleet ISI experience and the ISI experience specific to Waterford 3 is consistent with the Entergy evaluation that there is a low probability of surface-breaking flaws propagating due to fatigue.

Entergy indicated that fatigue is the only operative mechanism that could have caused flaws to either initiate or grow in the welds during the period since the previous inspection. The staff concludes that corrosion, stress-corrosion cracking, and other forms of degradation due to the material=s interaction with its chemical environment are not active degradation mechanisms for the RV welds. This is because the RV forgings and welds are separated from the reactor coolant by a layer of corrosion-resistant cladding. Even if the cladding was breached (for example, due to an original fabrication flaw in the cladding), the coolant water chemistry is controlled such that oxygen and other aggressive contaminants are maintained at very low levels so that the coolant is not aggressive to the ferritic material. Furthermore, the welds have not been subjected to a history of abnormal operational loading events, so mechanical overload has not been an active flaw initiation or propagation mechanism. Therefore, the staff agrees

with the conclusion that fatigue is the only likely operative mechanism that could have created or propagated flaws since the date the previous ISI examinations were performed.

The licensee states that the fatigue usage factors for these RPV welds will be much less than the ASME Code design limit of 1.0 after 40 years of operation, and that the most severe fatigue transient would be the cooldown event. The staff agrees that it is unlikely that more than one or two of these cooldown events would occur during the requested extension period of the licensee=s proposed alternative. In addition, the staff estimates that any flaw growth due to normal operational transients during the period since the last ISI examination would likely be very minimal.

Entergy provided the unirradiated nil-ductility transition reference temperature (RTNDT) values for each of the RV beltline materials for Waterford 3 and the PTS reference temperature (RTPTS) values to facilitate assessment of the effects of neutron irradiation on these beltline materials. Section 50.61 of 10 CFR currently provides PTS screening criteria of RTPTS equal to 270 degrees Fahrenheit (°F) for plates and axial welds and RTPTS equal to 300 °F for circumferential welds. Based on current projections, the licensee finds that the lower shell circumferential weld M-1004-2 is the most limiting RV beltline material for Waterford 3. The projected RTPTS value of 53 °F at 32 effective full-power years for this material is well below the applicable PTS screening criterion. Furthermore, it is recognized by the NRC and industry that a large amount of conservatism exists in the current PTS screening criteria as evidenced in the NRC memorandum, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Criteria in the PTS Rule (10 CFR 50.61)," dated December 31, 2002 (Reference 3).

Based on the beltline material data, existing PTS criteria, and the known conservatism of the current PTS rule, it is clear that Waterford 3 will remain well below the PTS screening criteria during the extension period. Therefore, the staff concludes that complying with 10 CFR 50.61 is sufficient to demonstrate that the probability of RV failure due to a PTS event is acceptably low.

The PTS risk associated with operation during any time interval is the product of the likelihood that a significant flaw exists and the likelihood that a PTS event occurs during the interval, which would challenge the flaw. An increased risk associated with the requested extension arises from the potential existence of a significant flaw that would have been detected and repaired during the inspection at the end of the original interval. With an extended interval, this flaw would continue to be vulnerable to a severe PTS event during the period the inspection interval is extended. Instead of attempting to estimate this increased risk, Entergy concluded that the low probability of a PTS event during the requested extension combined with the low probability of a flaw existing in the RV results in a very small probability of RV failure due to PTS.

Entergy identified the EOPs that would be used to respond to the three scenarios (developed by the NRC staff during its PTS risk reevaluation work) that are believed to be the most likely scenarios that could cause a PTS event, which would challenge significant flaws in the RV welds. The staff concurs that the likelihood of any of these initiating events occurring during the ISI interval extension period is low. Furthermore, existing plant procedures and material properties can mitigate the severity of, or the effects of, the PTS event that would be caused by these initiating events.

In summary, the staff has reviewed Entergy=s evaluation and makes the following conclusions:

1. Previous RV ISI results were of sufficient quality to provide useful test results.

2. Previous ISI examinations did not identify any indications.

3. The RV welds are not subjected to stresses or corrosive conditions that would create new flaws or cause old flaws to grow.

4. Industry experience with ISI examinations of similar welds has yielded similar results; there are no known significant RV weld flaws.

5. The most severe degradation mode that is expected to be operative is fatigue, and the most severe operational event with respect to fatigue is cooldown, an event expected to occur only one or two times during the requested extension period. Therefore, growth of flaws due to fatigue would be minimal during the period since the previous ISI examination and would likely be very small during the proposed extension period.

6. The RV material has sufficient toughness to be acceptable with respect to PTS, as determined by Entergy=s compliance with the requirements of 10 CFR 50.61.

7. The likelihood of a severe PTS event occurring during the proposed extension period is low.

Accordingly, the staff concurs with Entergy=s assessment that the Waterford 3 welds have a low likelihood of having significant flaws and experiencing a severe PTS event during the proposed second ISI interval extension period. The staff finds that the risk associated with the one-cycle extension of the ISI interval is sufficiently small that it need not be quantified to support the conclusion that this alternative continues to provide an acceptable level of quality and safety.

Operation of the RV for an additional cycle, without performing the ISI examination of the subject welds, would not significantly increase the risk of either flaw growth due to fatigue or RV failure due to PTS.

4.0 CONCLUSION

On the basis of the above evaluation, the NRC staff concludes that Entergy=s proposed alternative provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the staff authorizes the proposed alternative for the second 10-year ISI interval at Waterford 3. This proposed alternative is authorized until the end of the fall 2009 refueling outage (Refueling Outage RF16) for Waterford 3. All other requirements of the ASME Code for which relief has not been specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

5.0 REFERENCES

1. Entergy Operations, Inc., from F.G. Burford (Acting Director, Nuclear Safety and Licensing) to U.S. Nuclear Regulatory Commission Document Control Desk, "Request for Alternative W3-ISI-003 Proposed Alternative to Extend the Second 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations,"

dated April 26, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML071230119).

2. Letter from R. Gramm of the NRC to G. Bischoff of the Westinghouse Owners Group, "Summary of Teleconference with the Westinghouse Owners Group Regarding Potential One Cycle Relief of Reactor Pressure Vessel Shell Weld Inspections at Pressurized Water Reactors Related to WCAP-16168-NP,

'Risk-Informed Extension of Reactor Vessel In-Service Inspection Intervals,' "

dated January 27, 2005 (ADAMS Accession No. ML050250410).

3. NRC Memorandum, A. Thadani to S. Collins, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Criteria in the PTS Rule (10 CFR 50.61)," dated December 31, 2002 (ADAMS Accession No. ML030090629).

Principal Contributor: C. Sydnor Date: February 15, 2008