ML081710470

License Renewal Application, May 19, 2008 Draft Request for Additional Information, for Sections 4.2 and 4.4

ML081710470
Person / Time
Site:	Three Mile Island
Issue date:	05/19/2008
From:	Daily J NRC/NRR/ADRO/DLR
To:
Robinson, J NRR/DLR/RPB1 415-2878
Shared Package
ML081780009	List: ML081710470 ML081780006
References
TAC MD7701
Download: ML081710470 (6)

Text

DRAFT RAI SET dated 5/19/08 REQUEST FOR ADDITIONAL INFORMATION TMI UNIT 1 LICENSE RENEWAL APPLICATION RAI#: 4.2.0.0-01 LRA section: 4.2

Background:

In Section 4.2, "Neutron Embrittlement of the Reactor Vessel and Internals," of your License Renewal Application (LRA), neutron embrittlement is discussed in several contexts relating to fracture toughness of the reactor vessel and internals, and the impact on fluence values used in the analysis of the Reference Temperature for PTS (RTPTS) , Low-Temperature Over-Pressure Protection (LTOP),

Pressure-Temperature (PT) limit curves, and the Charpy Upper Shelf Energy (USE).

Issue: The submittal does not include a reference for the source of the fluence values used in the analysis of the RTPTS, LTOP, PT limit curves, and the Charpy USE. The text mentioned the NRC RVID2 data base, but these data do not constitute an acceptable reference for the current EOL values. In addition, the submittal concerning this analysis indicates a nonexistent document: a fluence analysis was prepared for TMI-1 that included a benchmark comparison to measured cavity dosimetry test results, and these projections were determined to meet the uncertainty requirements of Regulatory Guide 1.190, Revision 2.

Request:

1. Please submit the analysis mentioned above that was prepared for the EOL of extended operation for the staff to complete the review of Section 4.2.

2. Please note that Revision 2 for Regulatory Guide 1.190 does not exist. Please provide a corrected reference document.

Section 4.4 Leak Before Break Analysis of Primary System Piping Section 4.4.1 Fatigue Flaw Growth Analysis RAI#: 4.4.1.0-01 LRA section: 4.4.1

Background:

In Section 4.4.1, Fatigue Flaw Growth Analysis, the applicant states that it will use the metal fatigue of reactor coolant boundary aging management program (B.3.1.1) to monitor fatigue transient cycles and assure that the number of occurrences do not exceed design limits.

Issue:

It is not clear to the NRC staff exactly how the applicant will apply the metal fatigue aging management program (B.3.1.1) to monitor fatigue of the leak-before-break (LBB) piping.

Request:

1) Discuss the actions that will be taken, besides performing reanalysis of LBB components, if the fatigue transient cycles exceed design limits.

Enclosure 2

2) Discuss how often the metal fatigue aging management program monitors fatigue transient cycles.

3) The applicant states that only significant thermal and pressure transients are monitored. Discuss the definition of a significant thermal or pressure transient and provide the associated technical basis.

4) Identify the document that contains the definition of significant and insignificant transients.

RAI#: 4.4.1.0-02 LRA section: 4.4.1

Background:

In Section 4.4.1, Fatigue Flaw Growth Analysis, the applicant states that selected locations of the piping system are chosen for the fatigue crack growth analysis.

Issue:

The details of fatigue flaw growth analysis are not provided in Section 4.4.1. It is not clear whether the relevant information has been submitted to the NRC in previous licensing actions.

Request:

1) Identify specific locations that were chosen for the fatigue crack growth analysis.

2) Discuss the initial flaw size assumed in the fatigue analysis and the final flaw size.

3) The applicant states that the leak before break analysis is contained in topical reports BAW-1999 and BAW-1847. The staff has the BAW-1999 report but not the BAW-1847 report. Please submit a copy of the BAW-1847 report.

RAI#: 4.4.1.0-03 LRA section: 4.4.1

Background:

Recent industry experience has shown that Alloy 82/182 dissimilar metal welds are susceptible to primary water stress corrosion cracking (PWSCC). Industry and the NRC are currently working to resolve PWSCC in Alloy 82/182 welds with respect to the LBB analysis assumptions.

Issue:

Section 4.4.1 has not addressed the issue of potential PWSCC of Alloy 82/182 welds with respect to the LBB analysis assumptions for those pipes that have been approved for LBB technology.

Request:

1) Identify LBB piping that contain Alloy 82/182 dissimilar metal welds and identify the welds.

2) Discuss the actions that will be taken to mitigate and/or inspect the Alloy 82/182 welds in the LBB piping to ensure that primary stress corrosion cracking will not affect the structural integrity of the LBB piping.

3) Discuss the validity of the original LBB analyses in light of industry experience in primary stress corrosion cracking of Alloy 82/182 butt welds.

RAI#: 4.4.1.0-04 LRA section: 4.4.1

Background:

The NRC approved a 1.3% stretch power uprate for TMI-1 on July 28, 1988 (ADAMS ML003765237). The power uprate may change pressure and temperature of the primary coolant system which in turn may affect the structural integrity of the LBB piping.

Issue:

The applicant has not addressed the impact of the power uprate on the LBB piping and fatigue flaw growth analysis.

Request:

Discuss the impact of the operating conditions of power uprate on the LBB piping at the end of 60 years.

RAI#: 4.4.1.0-05 LRA section: 4.4.1

Background:

In Section 4.4.1, Fatigue Flaw Growth Analysis, the applicant discussed the fatigue flaw growth aspect of the LBB application as part of the TLAA. However, the TLAA should also include a history of the structural integrity of the subject LBB piping.

Issue:

It is not clear to the staff the inspection history of the LBB piping. The inspection history of the LBB piping will provide the NRC an understanding of how the structural integrity of the LBB piping may be managed during the extended period of operation.

Request:

Discuss the inspection history of the piping systems that have been approved for LBB, including inspection results and frequency.

Section 4.4.2 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings RAI#: 4.4.2.0-01 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, page 4-45, fourth paragraph, the applicant discusses an updated flaw stability analysis in support of a generic leak-before-break analysis. On page 4-46, third paragraph, the applicant states that a flaw stability analysis was performed using the lower-bound CASS fracture toughness curves. On page 4-46, fourth paragraph, the applicant discusses a revised analysis.

Issue:

It is not clear to the NRC staff the number of analyses that are discussed in Section 4.4.2 and whether the analyses have been submitted to the NRC.

Request:

1) Clarify whether these three analyses are the same analysis.

2) Provide the title and reference of the three analyses, if they are individual analysis.

3) Submit the analyses that have not been submitted previously.

RAI#: 4.4.2.0-02 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, page 4-46, fifth paragraph, the applicant states that Based on this analysis, it was determined that the TMI-1 RCP [reactor coolant pump]

CASS components meet all safety margin requirements Issue:

Because the applicant discusses several analyses in Section 4.4.2, it is not clear which analysis shows that the CASS components meet all safety margins.

Request:

Please clarify.

RAI#: 4.4.2.0-03 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, the applicant has not provided sufficient information regarding inspection of the RCP casing in general and CASS material in specific.

Issue:

The current ultrasonic testing (UT) technique has not been qualified to examine CASS material in accordance with the ASME Code,Section XI.

Therefore, the NRC seeks information regarding the inspection of the CASS material and associated Alloy 82/182 weld (if exists).

Request:

1) Discuss how the RCP casing, which is made of CASS material, can be examined to determine its structural integrity.

2) Discuss the inspection history of the RCP casing, including results.

3) If the welds between the RCP nozzles and the pipe are fabricated with Alloy 82/182 filler metal, discuss the inspection of the welds, including inspection history, results, future inspection frequency, and examination volume coverage.

RAI#: 4.4.2.0-04 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, the applicant discusses the analyses of the CASS material of RCP casing. However, the applicant has not discussed how the CASS material will be managed and monitored for potential degradation during the extend period of operation.

Issue:

It seems that Appendix B of the TMI-1 LRA does not contain an aging management program to manage thermal aging embrittlement of CASS components.

Request:

Discuss how the thermal aging embrittlement of the CASS materials in the LBB piping will be managed.

RAI#: 4.4.2.0-05 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, it is not clear to the NRC staff whether the applicant has addressed the staff position in CASS as specified in the letter below.

Issue:

By letter dated May 19, 2000, Christopher I. Grimes of the NRC forwarded Douglas J.

Walters of Nuclear Energy Institute an evaluation of thermal aging embrittlement of CASS components (ADAMS Accession ML003717179). In the NRCs evaluation, the NRC staff provided its positions on how to manage CASS components.

Request:

Discuss whether the CASS casing of the RCP satisfies the staff positions in its evaluation dated May 19, 2000.

RAI#: 4.4.2.0-06 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, the applicant has not identified all CASS materials in the LBB piping.

Issue:

It is not clear how many components that are made of CASS besides RCP casing. The current ultrasonic testing is not qualified by the ASME Code,Section XI, to examine CASS materials. Therefore, the NRC needs to understand the extent of the CASS material in the subject LBB piping.

Request:

In addition to the RCP casing, identify any components in the piping systems approved for LBB that are made of CASS material.

RAI#: 4.4.2.0-07 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Coolant Pump Casings, the applicant states that the lower-bound CASS material properties (e.g., fracture toughness) were used to show acceptability of CASS material for the period of extended operation.

Issue:

However, it is not clear in Section 4.4.2 whether the lower-bound CASS material properties are in fact bounding for the CASS material properties at the end of 60 years.

Request:

Please clarify.