ML051950261

American Society of Mechanical EngineersSection XI Inservice Inspection Program Relief Request to Extend the Second 10-year ISI Interval for Unit 1 Reactor Vessel Weld Examination - Request No. 1-ISI-27

ML051950261
Person / Time
Site:	Sequoyah
Issue date:	07/08/2005
From:	Pace P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1-ISI-27
Download: ML051950261 (20)

Text

Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 July 8, 2005 10 CFR 50.55a(a)(3)(i)

U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C. 20555-0001 Gentleman:

In the Matter of Docket No. 50-327 Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT (SQN) UNIT 1 -

SUBJECT:

AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

SECTION XI INSERVICE INSPECTION (ISI) PROGRAM RELIEF REQUEST TO EXTEND THE SECOND 10-YEAR ISI INTERVAL FOR UNIT 1 REACTOR VESSEL (RV)

WELD EXAMINATION -

REQUEST NO. 1-ISI-27

References:

1. Westinghouse Owners Group Topical Report, WCAP-16168-NP, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," October 2003

2. Letter from Nuclear Regulatory Commission to Westinghouse Electric

Company, "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 Pursuant to 10 CFR 50.55a(a)(3)(i), TVA hereby requests approval to use an alternative to the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, P-lted mecn Ad pow

U.S. Nuclear Regulatory Commission Page 2 July 8, 2005 Paragraph IWB-2412, Inspection Program B, for SQN Unit 1.

The alternative involves a proposed relief request to defer Unit 1 RV inspections for one fuel cycle.

The one-cycle deferral will allow additional time for completing evaluations and staff review associated with Westinghouse Owners Group Topical Report, WCAP-16168 (Reference 1).

The NRC has communicated to the Westinghouse Owners Group that the staff would agree to licensees submitting a one-cycle relief request for an extension.

It is also our understanding that TVA would have opportunity to request additional relief if review and approval of the topical report does not support our inspection interval.

SQN Unit 1 is currently in its second ISI interval.

The second interval began December 16, 1995, and considering ASME Code allowed extensions [see IWA-2430(d)], will end on May 31, 2006.

The total extension for both the first and second intervals for SQN Unit 1 is 209 days.

There are 167 days of this extension allocated to the second interval.

Accordingly, the second interval examination of the RV Shell Welds (Examination Category B-A) and the Nozzle-to-Vessel Welds and Inner Radius Sections (Examination Category B-D) is currently scheduled for April 2006, during the SQN Unit 1 Cycle 14 (UlC14) refueling outage.

Staff approval is requested to extend SQN's second inspection interval for the Examination Category B-A and B-D welds by one fuel cycle to allow examination during the Unit 1 Cycle 15 (UlC15) refueling outage (fall 2007).

The technical justification for this request is consistent with the guidance provided in reference 2. TVA's proposed extension of the ISI interval for these examinations will continue to provide an acceptable level of quality and safety, as described in the enclosed relief request.

TVA requests staff approval by February 2006 to support preparation activities for the April 2006 U1C14 refueling outage.

Nuclear Regulatory Commission Page 3 July 8, 2005 TVA's subject request is similar to other industry requests submitted by letters dated March 31, 2005 and June 8, 2005 for Palisades and Indian Point Nuclear Plants, respectively.

This letter contains no new commitments and no revisions to existing commitments.

If you have any questions about this change, please telephone me at (423) 843-7170 or J.D. Smith at (423) 843-6672.

Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosure cc (Enclosure):

Mr. Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop O-8G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739

ENCLOSURE TENNESSEE VALLEY AUTHORITY TVA SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

CODE, SECTION XI, Request for Relief 1-ISI-27 1.0 ASME Code Component(s) Affected The affected component is the SQN Unit 1 Reactor Vessel (RV),

specifically the following ASME Boiler and Pressure Vessel (BPV)

Code,Section XI examination categories and item numbers covering examinations of the reactor vessel.

These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Examination Item No Description Category B-A B1.11 Circumferential Shell Welds (W02-03, W03-04, W04-05, and W05-06)

B-A B1.21 Circumferential Head Welds (W01-02, and W09-10)

B-A B1.22 Meridional Shell Welds (W2A, W2B, W2C, W2D, W2E, and W2F)

B-A B1.30 Shell-to-Flange Weld (W06-07)

B-D B3.90 Nozzle-to-Vessel Welds (N11,

N12, N13, N14, N15, N16, N17, and N18)

B-D B3.100 Nozzle Inner Radius Areas (Nll-IR, N12-IR, N13-IR, N14-IR, N15-IR, N16-IR, N17-IR, and N18-IR)

(Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code Section XI is referred to as "the Code")

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2.0 Applicable Code Edition and Addenda

The SQN Unit 1 second Interval Inservice Inspection (ISI)

Program Plan is prepared to comply with the 1989 Edition of the Code.

3.0 Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100%

of reactor pressure vessel pressure-retaining welds identified in Table IWB-2500-1 once each ten-year interval.

In accordance with IWA-2430(d), the SQN Unit 1 second ISI interval is currently scheduled to conclude on May 31, 2006.

4.0 Reason for Request

An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of RV pressure retaining welds, Examination Categories B-A, and B-D welds, be performed once each ten-year interval.

Extension of the ISI interval for Examination Category B-A and B-D by one refueling cycle beyond the currently scheduled inspection is requested.

The intent of the requested one refueling cycle extension is to allow for deferment of the subject examinations to allow time for NRC review of industry efforts to extend the ISI interval for the subject examinations from 10 to 20 years.

These efforts use ASME Section XI Code Case N-691 (Reference 1) as a basis for using risk-informed insights to show that extending the inspection interval from 10 to 20 years results in a small change in the RV failure frequency that satisfies the requirements of Regulatory Guide 1.174 (Reference 2).

Following NRC approval of these efforts, TVA intends to submit a separate request to extend the current 10-year interval for SQN to 20 years.

The 20-year inspection interval will result in a reduction in man-rem exposure and examination costs.

5.0 Proposed Alternative and Basis for Use

The second ISI interval for SQN Unit 1 started on December 16, 1995 and will end on May 31, 2006.

SQN Unit 1 has used 209 days of the IWA-2430(d) allowed one-year extension.

In accordance with IWA-2430(d),

the subject examinations are currently scheduled to be performed during the spring 2006 refueling outage, (UlC14).

The proposed inspection date is one refueling E-2

outage beyond the Code required inspection interval, which would defer the subject examinations to the UlC15 refueling outage.

In accordance with 10 CFR 50.55a(a) (3) (i), this interval extension is requested on the basis that the current ISI interval can be extended while providing an acceptable level of quality and safety.

The requirements for a technical basis to extend the 10-year RV ISI interval by one refueling cycle are contained in a letter from R. Gramm of the NRC to G. Bischoff of the Westinghouse Owners Group, dated January 27, 2005 (Reference 3).

The technical justification for the extension was developed based on the guidance provided by the NRC in the January 27, 2005 letter.

The technical justification consists of five areas described below:

5.1 Plant specific reactor vessel inservice inspection history.

5.2 Fleetwide reactor vessel inservice inspection history.

5.3 Degradation mechanisms in the reactor vessel.

5.4 Material condition of the reactor vessel relative to embrittlement.

5.5 Operational experience relative to RV structural integrity challenging events.

5.1 Unit 1 Reactor Vessel Inservice Inspection History SQN Unit 1 is in its second ISI interval for the reactor pressure vessel examinations.

Therefore, the preservice inspections (PSI) and one ISI have been performed on the Examination Category B-A and B-D welds to date.

In summary, the PSI was performed in accordance with ASME Section XI, 1974 Edition and Summer 1975 Addenda; and ISIs have been performed in accordance with ASME Section XI and Regulatory Guide 1.150 (Reference 4).

The ISIs to date have achieved acceptable coverage (i.e., >90% or coverage approved in ISI program relief requests), and no reportable indications with the exception of RV head to head ring weld W09-10 found during the PSI.

This PSI flaw indication on weld W09-10 was originally detected and sized in accordance with the ASME Section XI, 1974 Edition with Addenda through Summer 1975, and found to exceed the allowable limits.

This flaw was required to meet the acceptability criteria of IWB-3600 and was subjected to successive examinations as required in IWB-2420(b) and (c).

The flaw was examined during the 1984 ISIs and was re-evaluated.

During this evaluation, the PSI and ISI data was reviewed and correlated.

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The 1984 evaluation of the flaw determined it to be allowable in accordance with IWB-3510 of the ASME Section XI 1977 Edition with Addenda through Summer 1978.

The flaw was again examined during the 1990 ISIs and the two examination results were reviewed.

The indication was then found to be essentially unchanged since the 1984 ISIs and was determined to be allowable in accordance with IWB-3510 of ASME Section XI, 1977 Edition with Addenda through Summer 1978.

The flaw was re-examined during the 1993 ISIs and recorded and sized with the guidelines of Regulatory Guide 1.150, Revision 1, Appendix A and the requirements of the 1986 Edition of ASME Section XI.

The data from the three previous examinations was also reviewed and compared.

The flaw was found to be within the allowable limits of Table IWB-3510-1.

In accordance with paragraph IWB-2420 (c),

the flaw has been examined for three successive inspection periods and remained essentially unchanged; therefore, the component examination schedule reverted to the original schedule of successive inspections.

Based on the examination method and coverage obtained, it is reasonable to conclude that the examinations were of sufficient quality to detect any significant flaws that would challenge RV integrity.

A most recent detailed inspection history for the welds to which the subject examinations apply is contained in the following table.

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TABLE SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently being being monitored*

monitored*

W02-03 B-A 4/27/1993 91% coverage None None None for reflectors oriented parallel to the weld.

67.4%

coverage for reflectors oriented transverse to the weld W03-04 B-A 5/4/1993 100 %

None None None coverage for reflectors oriented parallel to the weld.

100 %

coverage for reflectors oriented transverse to the weld W04-05 B-A 5/4/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld E-5

TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently being being monitored*

monitored*

W05-06 B-A 5/4/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld WO1-02 B-A 4/27/1993 64.2%

None None None coverage for reflectors oriented parallel to the weld.

61.5%

coverage for reflectors oriented transverse to the weld W09-10 B-A 2/3 of 98% coverage None None None weld for 5/2/1993 reflectors oriented 1/3 of parallel and weld transverse 3/9/1984 to the weld.

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TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently

• being being monitored*

monitored*

W2A B-A 4/30/1993 99.1%

None None None coverage for reflectors oriented parallel to the weld.

99% coverage for reflectors oriented transverse to the weld W2B B-A 4/30/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

94.7%

coverage for reflectors oriented transverse to the weld W2C B-A 4/30/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

99.9%

coverage for reflectors oriented transverse to the weld E-7

TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage 4 of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently being being monitored*

monitored*

(in)

W2D B-A 4/30/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

90.8%

coverage for reflectors oriented transverse to the weld.

W2E B-A 4/30/1993 97.3%

None None None coverage for reflectors oriented parallel to the weld.

99.9%

coverage for reflectors oriented transverse to the weld.

W2F B-A 4/30/1993 89.9%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld.

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TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently being being

.monitored* monitored*

____(in

)

W06-B-A 5/9/1993 85.3%

None None None 07 coverage for reflectors oriented parallel to weld.

64.8%

coverage for reflectors oriented transverse to the weld.

Ni1-B-D 5/9/1993 100%

None None None IR N12-B-D 5/9/1993 100%

None None None IR N13-B-D 5/9/1993 100%

None None None IR N14-B-D 5/9/1993 100%

None None None IR N15-B-D 5/9/1993 100%

None None None IR N16-B-D 5/9/1993 100%

None None None IR N17-B-D 5/9/1993 100%

None None None IR N18-B-D 5/9/1993 100%

None None None IR E-9

TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications Category indications currently currently being being monitored*

monitored*

(in)

Nil B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld N12 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld N13 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld E4 0

TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID Examination Inspected obtained reportable indications indications indications currently currently Category being being monitored*

monitored*

(in)

N14 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

100%

coverage for reflectors oriented transverse to the weld N15 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

54.8%

coverage for reflectors oriented transverse to the weld N16 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

38.8%

coverage for reflectors oriented transverse to the weld E41

TABLE (Continued)

SQN Unit 1 Inservice Inspection Results Weld ASME Date Last

% Coverage

1. of

1. of Growth of ID ExaminationC Inspected obtained reportable indications indications ategory indications currently currently being being monitored*

monitored*

(in)

N17 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

45.9%

coverage for reflectors oriented transverse to the weld N18 B-D 5/9/1993 100%

None None None coverage for reflectors oriented parallel to the weld.

37.2%

coverage for reflectors oriented transverse to the weld

• Note: Due to improvements in inspection technology, the most recent inspection is considered to be of the greatest quality of the inspections performed.

Therefore, the inspection data provided in this table is for the most recent inservice inspection.

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5.2 Fleetwide Reactor Vessel Inservice Inspection History As part of the technical basis for ASME Code Case N-691, a survey of RV ISI history for 14 pressurized water reactors was performed.

These 14 plants represented 301 total years of service and included RVs fabricated by various vendors.

The plants reported that no reportable findings had been discovered during examinations of Category B-A and B-D welds of their RVs.

It is widely recognized in the fracture mechanics community that fatigue crack growth of embedded flaws is substantially smaller than that of surface breaking flaws.

Surface breaking flaws in the RV cladding are typically a result of lack of fusion defects between bands of cladding.

In studies performed by Pacific Northwest National Laboratory for the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation, it was determined that in plants with multi-pass cladding, for a flaw to exist through the cladding, two flaws would have to be aligned on top of one another.

The probability of this occurring is very low

(<0.0001).

The SQN Unit 1 RV is constructed with multi-pass cladding and therefore has a low probability of containing through-cladding surface-breaking flaws.

All PWR plants have performed their first 10-year ISI of the subject examinations.

No surface-breaking or unacceptable near-surface flaws (i.e., defects) have been found in any of these inspections performed per the requirements of Regulatory Guide 1.150 or ASME Section XI, Appendix VIII.

5.3 Degradation Mechanisms in the Reactor Vessel The welds for which the subject examinations are conducted are similar metal low alloy steel welds.

The only currently known degradation mechanism for this type of weld is fatigue due to thermal and mechanical cycling from operational transients.

Studies have shown that while flaw growth of simulated flaws in a reactor vessel would be small, the operational transient which has the greatest contribution to flaw growth is the cooldown transient.

Based on operating experience, the cooldown transient is a low frequency transient and is not expected to occur more than a few instances during the requested inspection extension period.

Therefore, any flaw growth during the requested deferral period will be inherently small.

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The fatigue usage factors for the welds in the subject examinations are much less than the ASME Code design limit of 1.0 after 40 years of operation.

These usage factors are calculated using a very conservative design duty cycle.

It is very unlikely that more than a few of these events (e.g. heatup or cooldown) would actually occur during the extension period of this proposed alternative.

It is important to note that this request does not apply to any dissimilar metal welds, including Alloy 600 base metal or Alloy 82/182 weld material where primary water stress corrosion cracking is a concern or any other augmented inspection requirements imposed.

5.4 Material Condition of the Reactor Vessel Relative to Embrittlement The RV beltline is the limiting area in terms of embrittlement for the subject examinations.

The composition of each material in the RV beltline, along with fluence and embrittlement data, can be found in the NRC Reactor Vessel Integrity Database (RVID)

(Reference 5).

This information is provided for SQN Unit 1 in the following table.

Note: The RTPTS values have been updated as discussed below.

Sequoyah Unit 1 Material Values contained in the RVID Un-Irradiated Major Material Region Description RTuRTpTs Cu Ni P

@32 Type Heat Location

[wt%]

[wt%]

[wt%]

[OF]

Method EFPY 1

Forging 980807/

Intermediate 0.15 0.86 0.011 40.0 Plant 209 281489 Shell 05 Specific 2 Forging 980919/

Lower Shell 0.13 0.76 0.02 73.0 Plant 231 281587 04 Specific 3 Weld 25295 Circ Weld 0.35 0.13 0.021

-40.0 Plant 204 Specific

.~

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The 10 CFR 50.61 currently provides PTS screening criteria of RTPTs equal to 270'F for plates and axial welds and RTPTS equal to 300'F for circumferential welds.

Based on the current projections in WCAP-15293 (Reference 6), the lower shell forging is the limiting material for SQN Unit 1. The projected RTPTS value at 32 EFPY (2310F) for this material is well below the PTS screening criteria.

Furthermore, it is recognized by the NRC and industry that a large amount of conservatism exists in the current PTS screening criteria (Reference 7).

In the NRC PTS Risk Reevaluation, results have shown that it may be possible to remove an amount of conservatism equivalent to reducing a plant's RTPTS value by at least 700F.

While the exact amount of conservatism that will be removed has not been determined, it is clear that SQN Unit 1 will be below the current PTS screening criteria during the extension period and further below the potential revised PTS screening criteria.

5.5 Operational Experience Relative to RV Structural Integrity Challenging Events It is widely recognized that the greatest possible challenge to reactor pressure vessel integrity for a PWR is PTS.

A PTS event can be generally described as a rapid cooling of the RV followed by a late repressurization.

Plants have taken steps such as implementing emergency operating procedures (EOPs) and operator training to lower the likelihood of a PTS event occurring.

Due to the implementation of such measures, industry experience indicates the number of occurrences of PTS events fleetwide is very small.

When considered over the combined fleetwide PWR operating history, the frequency of PTS events is very small.

When considering the frequency of PTS events and the length of the requested extension, the probability of a PTS event occurring during the requested extension is also very low.

Combining the low probability of a PTS event with the low probability of a flaw existing in the RV (given the previously discussed inspection history), the probability of RV failure due to PTS is also very small.

SQN Unit 1 has implemented EOPs and operator training to prevent the occurrence of PTS events. Consistent with the Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGs), the SQN EOPs allow operators to identify the onset of PTS conditions and provide the steps required to mitigate any cold pressurization challenge to RV integrity.

The basic PTS mitigation strategy of the SQN EOPs involves 1) termination of the primary system cooldown, 2) termination of emergency core cooling system flow (if proper criteria are met), 3) depressurization of the primary system, 4) establishment of stable primary system conditions in E-5

the normal operating range, and 5) implementation of a thermal "soaking" period prior to any cooldown outside of the normal operating region.

By combining 1) the basic requirements of the WOG ERGs, 2) the use of plant specific setpoints with a defined technical

basis, and 3) the formal reconciliation of any differences between the WOG ERG reference plant and SQN, the SQN EOPs provide adequate means for preventing potential PTS transients.

The current requirements for inspection of RV pressure-retaining welds have been in effect since the 1989 Edition of ASME Code,Section XI.

The industry has expended significant cost and man-rem exposure that have shown no service-induced flaws in the ASME Section XI Examination Category B-A or B-D RV welds.

ASME Section XI Code Case N-691 and industry efforts have shown that risk insights can be used to extend the reactor vessel ISI interval from 10 to 20 years.

The 10-year extension satisfies the change in risk requirements of Regulatory Guide 1.174; and, in accordance with 10 CFR 50.55a(3)(i), maintains an acceptable level of quality and safety.

Based on these efforts having shown that the risk of vessel failure with a 10-year inspection interval extension is low and achieves an acceptable level of quality and safety, it is reasonable to conclude that one refueling cycle extension will also achieve an acceptable level of quality and safety.

On the basis of the above discussion, the risk associated with extending the inspection interval by one refueling cycle is small.

Therefore, TVA considers the proposed alternative for the subject examinations at SQN Unit 1 to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(3)(i).

6.0 Duration of Proposed Alternative

The alternative is requested to extend the second ISI interval for one refueling cycle for the Examination Category B-A and B-D RV welds beyond the ASME Code required 10-year inspection interval and the Code allowed 167-day extension.

This request is applicable to the second inspection interval only.

If this relief request is approved, the second ISI interval will end at the conclusion of the fall 2007 UlC15 outage for the subject examinations.

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7.0 References

1.

ASME Boiler and Pressure Vessel Code, Code Case N-691, "Application of Risk-Informed Insights to Increase the Inspection Interval for Pressurized Water Reactor Vessels,"

Section XI, Division 1, November 2003.

2.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.

3.

R. Gramm of the NRC to G. Bischoff of the WOG, "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.

4.

Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds during Preservice and Inservice Examinations,"

February 1983.

5.

Nuclear Regulatory Commission Reactor Vessel Integrity Database, Version 2.0.1, dated July 6, 2000.

6.

WCAP-15293, Revision 02, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," July 2003.

7.

NRC Memorandum, Thadani to 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.

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