ML082560178

Response to NRC Request for Additional Information Related to Three Mile Island Nuclear Station, Unit 1, License Renewal Application

ML082560178
Person / Time
Site:	Three Mile Island
Issue date:	09/10/2008
From:	Gallagher M P AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-08-20164, TAC MD7701
Download: ML082560178 (34)

Text

AmerGen.S Michael P. GallagheT, PE Vice President License Renewal Projects AmerGen 200 Exelon Way KSA/2-E Kennett Square, PA 19348 Telephone 610.765.5958 www.exeloncorp.com michaelp.gallagher@exeloncorp.com An Exelon Company 10 CFR 50 10 CFR 54 5928-08-20164 September 10, 2008 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Three Mile Island Nuclear Station, Unit 1.Facility Operating License No. DPR-50 NRC Docket No.50-289

Reference:

Response to NRC Request for Additional Information related to Three Mile Island Nuclear Station, Unit 1, License Renewal Application Letter from Mr. Jay Robinson (USNRC), to Mr. Michael P. Gallagher (AmerGen)"Request for additional information for sections 4.2 and 4.4 of the Three Mile Island Nuclear Station, Unit 1, License Renewal Application", dated August 2 0 th, 2008. (TAC No. MD7701)In the referenced letter, the NRC requested additional information related to Sections 4.2 and 4.4 of the Three Mile Island Nuclear Station, Unit 1, License Renewal Application (LRA).Enclosed are the responses to this request for additional information.

In support of our responses, two reports are also enclosed.This letter and its enclosure contain no commitments.

If you have any questions, please contact Fred Polaski, Manager License Renewal, at 610-765-5935.

I declare under penalty of perjury that the foregoing is true and correct.Respectfully, Executed on 40- /a '2o040 Michael P. GallagherI Vice President, License Renewal AmerGen Energy Company, LLC September 10, 2008 Page 2 of 2 5928-08-20164 Enclosure A: Response to Request for Additional Information for sections 4.2 and 4.4 of the Three Mile Island Nuclear Station, Unit 1, license renewal application.

Attachments:

1. AREVA Report 86-9038511-000, "TM I License Renewal RPV Final Fluence Report," dated March 15, 2007 (RAI # 4.2.0.0-01)

2. Framatome Technologies Report 51-5000709-00,"Assessment of TLAA Issues in LBB Analysis of RCS Primary Piping," dated 1/30/98 (RAI #: 4.4.2.0-01) cc: Regional Administrator, USNRC Region I, w/ Enclosure (w/o attachments)

USNRC Project Manager, NRR -License Renewal, Safety, w/Enclosure USNRC Project Manager, NRR -License Renewal, Environmental, w/o Enclosure USNRC Project Manager, NRR -TMIGS, w/o Enclosure USNRC Senior Resident Inspector, TMIGS, w/o Enclosure File No. 08001 Enclosure

-A September 10, 2008 Page 1 of 15 5928-08-20164 Enclosure

-A Response to Request for Additional Information for sections 4.2 and 4.4 of the Three Mile Island Nuclear Station, Unit 1, License Renewal Application.

Enclosure

-A September 10, 2008 Page 2 of 15 5928-08-20164 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 Pressurized Thermal Shock (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 Reactor Vessel Integrity Database, Revision 2 (RVID2), but these data do not constitute an acceptable reference for the current end of life (EOL) values. In addition, the submittal concerning this analysis refers to 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 data 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.AmerGen Response 1) AREVA Report 86-9038511-000, "TMI License Renewal RPV Final Fluence Report," dated March 15, 2007 is the fluence analysis developed in support of TMI-1 License Renewal using the methodology described in BAW-2241 NP-A, which was previously reviewed and approved by the NRC. This fluence analysis is submitted for your review as Attachment 1 to this document.2) The reference to Regulatory Guide 1.190, Revision 2 was an error. The correct reference is Regulatory Guide 1.190, dated March 2001.

Enclosure

-A September 10, 2008 Page 3 of 15 5928-08-20164 RAI#: 4.4.1.0-01 LRA section: 4.4.1

Background:

In Section 4.4.1, Fatigque 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 how often the metal fatigue aging management program monitors fatigue transient cycles.2) 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.Amergen Response: The Metal Fatigue of Reactor Coolant Pressure Boundary aging management program B.3.1.1 is applicable to all Reactor Coolant System (RCS) piping and components.

It is used to assure that the numbers of actual plant transients do not exceed the numbers of transients used as inputs in the design fatigue analyses for these components.

The Leak-Before-Break analysis is applicable only to the large-bore RCS piping and the Reactor Coolant Pumps; therefore these components are all within the scope of the monitoring program. The LBB analysis used the same numbers of design transients as the design fatigue analyses since they were derived from the same functional specifications.

Therefore, the monitoring program will also assure that the numbers of cycles used in the LBB analysis will not be exceeded during the period of extended operation.

1) When a plant transient occurs, control room operators are required to document the event in the Transient Cycle Log if the transient meets any of the transient definitions provided in the TMI-1 fatigue monitoring procedure.

They also record thermal, pressure, flow, level and/or actuation data as required for the particular transient that occurred to enable validation of the transient type. The fatigue monitoring program engineer is required to review the transient logbook semi-annually, validate that each actual transient is bounded by the applicable design transient definition, update the cycle counts, compare actual numbers of cycles to limits, and prepare a transient summary report. Corrective actions are triggered when it is recognized that a transient type is approaching 80 percent of its limit.2) Transients are deemed "significant" if they affect stress cycles significantly due to the rate of change of RCS temperature and pressure during the event. Transients can be divided into two main categories:

trip and non-trip.

The primary difference in these two Enclosure

-A September 10, 2008 Page 4 of 15 5928-08-20164 categories with regard to stress cycles is the rate of change of RCS temperatures and pressures.

The reactor trips exhibit much faster changes in RCS temperature than the non-trip transients (approximately ten times faster), and are therefore monitored.

Non-trip transients are also considered significant if they result in a high rate of change of core average temperature.

Examples include heatups and cooldowns (450-degree change at a rate between 0.5 and 1.5 deg/minute), Integrated Control System runbacks (up to 10 deg/minute), and 10% step changes, etc. For monitoring purposes, "non-significant" transients have a negligible impact on stress cycles. "Non-significant" transients include those that meet all five of the following criteria.

The transient:

a. Does not result in a power change of >10% in less than 2 minutes; and b. Does not involve an RC System or Component or Steam Generator test involving a flow, temperature or pressure change; and c. Does not result in a change of >10 deg F in RCS Cold Leg or Hot Leg temperature; and d. Does not result in initiation of HPI or AFW flow or in the opening of a PORV; and e. Does not result in a change of >25 deg F in Feedwater temperature at a rate >1 deg F/sec.

Enclosure

-A September 10, 2008 Page 5 of 15 5928-08-20164 RAI# 4.4.1.0-02:

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, Fatigue Flaw Growth Analysis, 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.AmerGen Response: 1) Within the scope of the Leak-Before-Break (LBB) analysis, there are a total of eight Alloy 82/182 dissimilar metal welds that are associated with the suction and discharge nozzles of the four Reactor Coolant Pumps (RCP). A forged stainless steel safe end pipe is installed in each line between the Cast Austenitic Stainless Steel (CASS) RCP nozzle and the carbon steel pipe. The Alloy 82/182 welds join the forged stainless steel safe end pipe to the carbon steel RCS piping.2) Currently, the Alloy 82/182 dissimilar metal welds are scheduled for examination every third refueling outage per the guidance of MRP-1 39, "EPRI Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines." The initial MRP-1 39 examinations are required to be completed no later than December 31, 2010.This includes UT and bare metal visual inspection as specified in MRP-139, Tables 6-1 and 6-2. MRP-1 39, Table 6-1 (page 6-12), PWSCC Category E, is appropriate for these Alloy 82/182 welds and specifies the volumetric inspection requirement as once every 6 years. MRP-139, Table 6-2 (page 6-14) PWSCC Category K, specifies the frequency for visual inspections.

Any future mitigation actions would be in accordance with requirements of 10 CFR 50.55a and ASME Section XI.3) MRP-140, "Materials Reliability Program: Leak-Before-Break Evaluation for PWR Alloy 82/182 Welds" evaluates the impact of PWSCC of Alloy 82/182 welds on the LBB analysis.

Section 10, Summary and Conclusions, provides the following information: "The only change from existing LBB evaluations that needs to be addressed for Alloy Enclosure

-A September 10, 2008 Page 6 of 15 5928-08-20164 82/182 locations is consideration of PWSCC in these alloys. In this respect, three issues need to be revisited:

-Will the presence of PWSCC assure that cracks will grow in the through-wall direction before growing in the circumferential direction such that crack profiles consistent with the Duane Arnold Energy Center (DAEC) safe end intergranular stress corrosion cracking (IGSCC) pattern in the early 1980s will not result?-Will the morphology associated with PWSCC have a significant impact on the leakage rate calculation (or leakage flaw sizes) so as to affect the LBB margins and conclusions?

-Will the crack growth associated with PWSCC in Alloy 82/182 materials affect leak detection before the critical flaw size is reached?" MRP-140 also provides the answers to these questions as follows: "- An evaluation was performed to determine the growth direction of flaws associated with PWSCC. It was determined that based on both experimental studies and field behavior that flaws resulting from PWSCC will most likely grow in the through-wall direction and result in leak before growing in the length direction thus assuring LBB.-Even though the PWSCC morphology has an effect on the leakage rate calculation by increasing the leakage flaw size, adequate margins are still maintained between the critical flaw size and the leakage flaw size for all piping.-The time to grow a flaw from a leakage flaw to a critical flaw size is on the order of years. This shows that there is adequate time to detect leaks without concern for imminent pipe rupture." MRP-140 concludes that: "there is no concern for LBB applied to Alloy-82/182 locations in PWRs. The main loop piping has critical-to-leakage flaw size margins of at least two when PWSCC morphology is considered and therefore qualifies for LBB."

Enclosure

-A September 10, 2008 Page 7 of 15 5928-08-20164 RAI#: 4.4.1.0-03 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: 1) Discuss the impact of the operating conditions of power uprate on the LBB piping at the end of 60 years.AmerGen Response: 1) The LBB analyses applicable for TMI-1 were developed using the reactor coolant system design operating temperatures and pressures stated in Chapter 3 of the TMI-1 FSAR that are based upon 2568 MWt, which is both the design power level and licensed power level after the 1.3 percent power uprate. Therefore, there is no impact from the 1.3%stretch power uprate.Even though the design power level for TMI-1 is 2568 MWt, TMI-1 was initially licensed to 2335 MWt on the basis of the original design parameters for the main turbine-generator.

Subsequent modifications were made to the turbine blading that resulted in significant improvements in turbine efficiency and therefore plant electrical output. As a result, GPUN requested a license amendment to increase the licensed power level to the design power level of 2568 MWt.The staff issued a Safety Evaluation approving License Amendment No. 143 on July 26, 1988. This SER explained that the staff evaluated the fuel system design, the nuclear design, thermal hydraulic design, accident and transient analysis and Technical Specification changes for Cycle 7 and concluded that the proposed power uprate does not change the original design conditions and that all existing reactor design and safety criteria are preserved at the higher power level of 2568 MWt. With this 1.3% stretch power uprate from 2335 MWt to 2568 MWt, TMI-1 is operating at the original design conditions.

Enclosure

-A September 10, 2008 Page 8 of 15 5928-08-20164 RAI#: 4.4.1.0-04 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 time-limited aging analysis (TLAA). However, the TLAA should also include a history of the structural integrity of the subject LBB piping.Issue: The inspection history of the LBB piping is not clear to the staff. The inspection history of the LBB piping will provide the NRC with an understanding of how the structural integrity of the LBB piping may be managed during the extended period of operation.

Request: 1) Discuss the inspection history of the piping systems that have been approved for LBB, including inspection results and frequency.

AmerGen Response: 1) The TMI-1 LBB analysis evaluates the large-bore Reactor Coolant System (RCS) piping and the Reactor Coolant Pump (RCP) casings. This includes the 36-inch hot leg piping that connects the reactor vessel to the steam generators, the 28-inch cold leg piping that connects the steam generators to the reactor coolant pumps, and the 28-inch cold leg piping that connects the reactor coolant pumps to the reactor vessel. These components are subject to periodic examination by the ASME Section Xl program, as described below.The 36-inch diameter carbon steel hot leg piping includes a total of 36 Category B-J (pressure-retaining) welds (24 circumferential and 12 longitudinal).

The 28-inch diameter carbon steel cold leg piping includes a total of 119 Category B-J welds (71 circumferential and 48 longitudinal) and 8 Category B-F welds (Alloy 600 welds that connect the carbon steel cold leg piping to the forged stainless steel safe ends attached to the Reactor Coolant Pump nozzles).From original plant startup in 1974 through the 2003 refueling outage, the traditional ASME Section Xl program was used. This includes all three periods of the first and second ten-year inspection intervals and the first period of the third ten-year inspection interval.

The TMI-1 Inservice Inspection (ISI) Program required 100 percent of the 8 Category B-F welds to be surface-examined and volumetrically examined during each ten-year inspection interval.

Each of these welds was examined during each of the first two 10-year inspection intervals with satisfactory results. The ISI program also required 25 percent of the combined total of Category B-J and Category B-F circumferential welds to be examined in accordance with ASME Section Xl or alternatives approved by the NRC during each Ten-Year Inspection Interval.

The required sample of these welds was examined during the first two ten-year inspection intervals with acceptable results.During the first period of the third ten-year inspection interval, additional examinations were performed for six circumferential Category B-J welds, one longitudinal Category B-Enclosure

-A September 10, 2008 Page 9 of 15 5928-08-20164 J weld, and one Category B-F (Alloy 600) weld within the LBB piping. Each examination had acceptable results.Beginning with the second period of the third inspection interval, the TMI-1 ISI program was changed to a risk-informed program. The NRC approved the TMI-1 Risk-Informed Inservice Inspection (RISI) Program under relief request RR-00-21 in November 2003. It was implemented for examinations in the second period of the third ten-year inspection interval, which began with the 2005 refueling outage. The RISI program characterizes the previous Category B-J and Category B-F welds as Category R-A, Medium Risk Category 4 welds. The RISI program requires examination of 10 percent of the total population of the RCS Medium Risk Category 4 welds during each ten-year inspection interval.

No examinations have been completed for these LBB welds under the RISI program to-date.Additionally, the eight Category B-F (Alloy 600) welds are subject to minimum examination requirements from MRP-139, "EPRI Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines (MRP-139)." The initial MRP-139 volumetric examinations are required to be completed no later than December 31, 2010. Subsequent volumetric and bare metal visual examinations are performed as specified in Tables 6-1 and 6-2 of MRP-139. Table 6-1 (page 6-12), PWSCC Category E, is appropriate for these Alloy 82/182 welds and it specifies the volumetric inspection requirement as once every 6 years. Table 6-2 (page 6-14), PWSCC Category K, specifies the frequency for visual inspections as once every three refueling outages.The TMI-1 ISI program specifies examinations of these Alloy 600 welds in accordance with these MRP-1 39 requirements.

The Reactor Coolant Pump RC-P-1 B interior surface was visually examined during the 1981-1984 outage in accordance with ASME Section Xl requirements due to a maintenance disassembly and the results were satisfactory.

The Reactor Coolant Pump RC-P-1C interior surface was visually examined in accordance with ASME Section Xl requirements during the 1999 refueling outage due to a maintenance disassembly and the results were satisfactory.

Enclosure

-A September 10, 2008 Page 10 of 15 5928-08-20164 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.

On 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: The number of analyses that are discussed in Section 4.4.2 and whether the analyses have been submitted to the NRC and whether or not the CASS components meet all safety margins is not clear to the NRC staff.Request: 1) Clarify whether these analyses are the same analysis.2) Provide the title and reference of the analyses, if they are individual analyses.3) Submit the analyses that have not been submitted previously to the NRC.AmerGen Response: 1) Each of the three paragraphs in LRA Section 4.4.2 refer to the same updated flaw stability analysis.

The original flaw stability analysis applicable for TMI-1 is described in BAW-1999, "TMI-1 Nuclear Power Plant Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping," April 1987, Sections 3.2.3 and 4.5. This analysis was performed in accordance with the criteria specified in NUREG-1061, Volume 3. The updated flaw stability analysis was issued in 1998 to address thermal aging of the CASS RCP casings for the period of extended operation in accordance with the criteria specified in the Standard Review Plan 3.6.3. This updated flaw stability analysis is described in Framatome Technologies Report 51-5000709-00, "Assessment of TLAA Issues in LBB Analysis of RCS Primary Piping," dated 1/30/98, Section 4.2.5,"Evaluation of CASS Pump Casing Nozzles." 2) Framatome Technologies Report 51-5000709-00, "Assessment of TLAA Issues in LBB Analysis of RCS Primary Piping," dated 1/30/98.3) Report 51-5000709-00 is submitted as Attachment

2.

Enclosure

-A September 10, 2008 Page 11 of 15 5928-08-20164 RAI#: 4.4.2.0-02 LRA section: 4.4.2

Background:

In Section 4.4.2, Thermal Aqinq Embrittlement of Cast Austenitic Stainless Steel (CASS)Reactor Coolant Pump Casinqs, the applicant has not provided sufficient information regarding inspection of the reactor coolant pump (RCP) casing in general and CASS material in specific.Issue: The current ultrasonic testing (UT) technique has not been qualified through performance demonstration to thoroughly 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 methods and frequency, and examination volume coverage.AmerGen Response: LRA Table 3.1.2-1 provides the Aging Management Review Results for the Reactor Coolant System, and on page 3.1-64 provides the results for the Reactor Coolant Pump Casing. One of the line items identifies Loss of Fracture Toughness due to Thermal Aging Embrittlement as an Aging Effect Requiring Management, and identifies the Aging Management Program as the ASME Section Xl, Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1) program.Appendix B, Section B.2.1.1 further describes this program. The following responses address Request items 1, 2, and 3: 1) The Three Mile Island Nuclear Station Unit 1 Inservice Inspection (ISI) Program Plan, Third Ten-Year Inspection Interval, invokes the inspection requirements from the 1995 Edition, 1996 Addenda of ASME Section Xl for ASME Class 1 components.

Table IWB-2500-1 of ASME Section Xl categorizes pump casings as Examination Category B-L-2. Visual, VT-3 examination of the internal surface is required only when a pump is disassembled for maintenance, repair, or volumetric examination.

In accordance with these requirements, Inservice Inspection Summary Table 7.1-1 of the ISI Program Plan also specifies Visual VT-3 examination of Examination Category B-L-2 Pump Casings. The TMI-1 Reactor Coolant Pump casings do not contain Category B-L-1 welds, therefore no volumetric examination of the pump casings is required by ASME Section Xl or by the ISI Program Plan.2) The TMI-1 reactor coolant system has a total of four reactor coolant pumps; RC-P-1A, RC-P-1 B, RC-P-1C, and RC-P-1D. Pump RC-P-1B was visually examined during the 1981 -1984 outage in accordance with ASME Section Xl requirements Enclosure

-A September 10, 2008 Page 12 of 15 5928-08-20164 due to a maintenance disassembly and the results were satisfactory.

Pump RC-P-1C was visually examined during the 1999 refueling outage due to a maintenance disassembly and the results were satisfactory.

3) A forged stainless steel safe end separates each CASS RCP nozzle from the carbon steel RCS piping. A stainless steel weld joins each CASS RCP nozzle to the forged stainless steel safe end pipe. Therefore, there are no Alloy 82/182 welds joining the CASS pump casing nozzles to the pipe.

Enclosure

-A September 10, 2008 Page 13 of 15 5928-08-20164 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, 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 NRC's evaluation, the NRC staff provided its positions on how to manage CASS components.

Request: Discuss how the CASS casing of the RCP satisfies the staff positions in its evaluation dated May 19, 2000.Amergen Response: The RCP casing satisfies the staff positions in its evaluation dated May 19, 2000. The NRC evaluation, under Valve Bodies and Pump Casings, page 9, states: "Valve bodies and pump casings are adequately covered by existing inspection requirements in Section Xl of the ASME Code, including the alternative requirements of ASME Code Case N-481 for pump casings.Screening for susceptibility to thermal aging is not required and the current ASME Code inspection requirements are sufficient." In addition, Table 3 of the NRC evaluation (page 8)specifies ASME Section Xl examination requirements for CASS Pump Casings (Base Metal).The Three Mile Island Nuclear Station Unit 1 ISI Program Plan, Third Ten-Year Inspection Interval, invokes the inspection requirements from the 1995 Edition, 1996 Addenda of ASME Section Xl for ASME Class 1 components.

Table IWB-2500-1 of ASME Section Xl categorizes pump casings as Examination Category B-L-2. VT-3 visual examination of the internal surface is required only when a pump is disassembled for maintenance, repair, or volumetric examination.

In accordance with these requirements, Inservice Inspection Summary Table 7.1-1 of the ISI Program Plan also specifies Visual, VT-3 examination of Examination Category B-L-2 Pump Casings.The TMI-1 Reactor Coolant Pump casings do not contain Category B-L-1 welds; therefore, no volumetric examination of the pump casings is required by ASME Section XI.

Enclosure

-A September 10, 2008 Page 14 of 15 5928-08-20164 RAI#: 4.4.2.0-04 LRA Section: 4.4.2

Background:

In Section 4.4.2, Thermal Aqingq Embrittlement of Cast Austenitic Stainless Steel (CASS)Reactor Coolant Pump Casingqs, the applicant has not identified all CASS materials in the LBB piping.Issue: It is not clear how many components are made of CASS besides the RCP casing. The current ultrasonic testing has not been qualified through performance demonstration in accordance with the ASME Code,Section XI, Appendix VIII, 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.AmerGen Response: The Leak-Before-Break analysis for TMI-1 is applicable to the main Reactor Coolant System (RCS) piping. The Reactor Coolant Pump (RCP) casings are the only CASS components within the TMI-1 RCS primary piping, and therefore the only CASS components within the scope of the LBB analysis.

Enclosure

-A September 10, 2008 Page 15 of 15 5928-08-20164 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, 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.AmerGen Response: The lower-bound Charpy-impact energy and fracture toughness properties for the CASS pump casings described in Section 4.4.2 will not reduce further over time, and are therefore bounding for the CASS material properties at the end of 60 years. This is because the property values were developed from lower-bound fracture toughness curves prepared by Framatome Technologies in accordance with NUREG/CR-6177, "Assessment of Thermal Embrittlement of Cast Stainless Steels," May 1994.NUREG/CR-6177 provides two methods for predicting Charpy-impact energy and fracture toughness values of CASS, as described below.The first method estimates the extent of thermal embrittlement at saturation, i.e., the minimum impact energy that can be achieved for the material after long-term aging, and is determined based upon actual values for the chemical composition of the steel.The second method, which is the lower-bound method, provides an even more conservative estimate of the fracture toughness values when specific chemical composition of the CASS material is unknown. A predicted lower-bound J-R curve is developed for cast stainless steels of unknown chemical composition for a given grade of steel, ferrite content, and temperature.

The lower-bound curve is based upon the worst-case material condition and also produces values that will not reduce further over time.Framatome elected to use this second method in determining the fracture toughness values described in Section 4.4.2 because it is simpler and provides satisfactory results. Therefore, the analysis described in Section 4.4.2 developed predicted lower-bound fracture toughness values that are bounding for the CASS material properties at the end of 60 years.

Enclosure A, Attachment

1. 2 AMERGEN LETTER # 5928-08-20164 17 PAGES TOTAL 20032-22 (12/96)RELEASE /DOCUMENT RELEASE NOTICE DATE4//f FRAMATOME (DRN) RECORDS TC HH No LO 0 1E5 MGMT. INITIALS CONTRACT NUMBER PLANT CC OR CHARGE NUMBER PAGE 4110102 ANO-1, ONS-1,2,&

3, TMI-1 41020 1 OF PUL COM PART NO. OR TASK NO. F rG DOCUMENT DOCUMENT TITLE SAFETY STAT ATT?(NIA if not applicable)

NUMBER (30 CHARACTER MAXIMUM) CLASSIF. (YIN) (YIN)51 -5000709 -00 ASSESS. OF TLAA ISSUES IN LBB. S N N REQUIRED DISTRIBUTION NO. COPIES INFORMATION COPIES NO. COPIES (ORGANIZATION OR TE) DRN DOC COM (INDIVIDUAL'S NAME) DRN DOC COM D. J. FIRTH (OF-57) 1 MKMAYO (OF-50) +ORIG. 1 -M. A. RINCKEL (OF-54) 1 1 S. FYFITCH 1 4 A. D. NANA (OF-50) 1 1 -K. K. YOON 1 1 PROJECT MANAGEMENT FUNCTION APPROVAL: (SIGNATURE)

' -.____________(PRINTED NAME) *7D: J g,~-~ DATE RELEASED BY (PREPARER):

I (PRINTED NAME) A. D. N DATE A. A. RINCKEL DATE I PROCEDURE 0412-66 20440-7 (12/95)mRAMATOME TECH NOLOG IES ENGINEERING INFORMATION RECORD Document Identifier 51 -5000709-00 Title ASSESSMENT OF TLAA ISSUES IN LBB ANALYSIS OF RCS PRIMARY PIPING PREPARED BY: REVIEWED BY: Name A. D. NANA Name M. A. RINOKEL Signature

.Date 12/31/97 Technical Manager Statement:

Initials Signature Date 12/31/97 Reviewer is Independent.

Remarks: This document assesses the adequacy of the existing LBB analysis of the RCS primary piping, for the period of extended operation, for the utilities participating in the B&WOG Generic License Renewal Program (GLRP). The nuclear plants of the participating utilities are ANO-1, ONS-1, ONS-2, ONS-3, and TMI-1. The portions of the analyses that are considered to be time limited aging analysis (TLAA) are: a) fatigue flaw growth analysis and, b) thermal aging.The above two TLAA issues are addressed in this document.

The results of the assessment is given in Section 5.0. Based on this assessment, the following conclusions are made.a) Fatigue flaw growth analysis is not required for the period of extended operation as a result of an issue of new S.R.P.3.6.3 (Reference 9).b) Thermal aging of CASS RCP nozzles should not be a concern for the period of extended plant operation based on the results of the flaw stability analysis of Reference 17.Page 1 of 16

'r FRAC ATOM E Doc. I.D. 51-5000709-00 CONTENTS 1. IN T R O D U C T IO N ................................................................................................................

3........

3 2. DEVELOPMENT OF LBB REQUIREMENTS

.................................................................

4 2 .1. E vo lutio n of L B B .......................................

............................................................................

4 2.2. M odifications to G D C -4 ......................................................................................................

.4 2.3. Development of New SRP for LBB ................................................................................

4 3. FATIGUE FLAW GROWTH ANALYSIS -TLAA ISSUE 1 .......................

6 3.1. LBB Analysis of RCS Primary Piping .............................................................................

6 3.2. Acceptance Criteria for LBB Submittals

.......................................................................

6 3.3. Elimination of Requirements for Fatigue Crack Growth Analysis ...............................

7 4. THERMAL AGING -TLAA ISSUE 2 .................................................................................

8 4 .1 .M a in C o o la nt P ip in g ................................................................................................................

8 4.2. Cast Austenitic Steel (CASS) Components

...............................

8 4.2.1. Justification in LBB Submittal

........................................................................................

9 4.2.2. R ecent D evelopm ent ....................................................................................................

..9 4.2.3. Assessment of Fracture Toughness Properties of CASS Materials

...................

10 4.2.4. Approach for Evaluation of CASS Pump Casings ...................................................

12 4.2.5. Evaluation of CASS Pump Casing Nozzles ..............................................................

12 5 .S U M M A R Y .................................................................................................................................

14 6 .R E F E R E N C E S ..........................................................................................................................

15 Page 2 TRItO CH N ,MLe, Doe. I.D. 51-5000709-00

1. INTRODUCTION The LBB analyses for the RCS primary piping of the B&W Designed NSS (References 1 and 2) have been reviewed and approved by the NRC for the current licensing period. These reports have successfully demonstrated the application of "leak-before-break" to the RCS primary piping system.This document assesses the adequacy of the existing LBB analyses (References 1 and 2 with supporting calculation packages), for the period of extended operation, for the utilities participating in the B&WOG Generic License Renewal Program (GLRP). The nuclear plants of the participating utilities are ANO-1, ONS-1, ONS-2, ONS-3, and TMI-1. The LBB evaluations included fatigue flaw growth, flaw stability, and limit load analyses.

In addition, the reports of the LBB analysis qualitatively addressed the thermal aging issue relative to the RCS primary piping and the RC Pump casings for the current licensing period.The portions of the analyses that are considered to be time limited aging analysis (TLAA) require to be assessed for the period of extended operation.

These are: 1) fatigue flaw growth analysis and, 2) thermal aging.The above two TLAA issues are addressed in Sections 3 and 4 respectively.

The historical development of the regulations involved in the application of LBB technology to high energy piping in the nuclear power plants is provided in Section 2, to help clarify the discussions in Section 3. A summary of the assessment of the TLAA issues in the RCS primary piping LBB analyses is given in Section 5.Page 3 FR IC1NAMA00OMS Doc. I.D. 51-5000709-00 TECH NO LOG I S 2. DEVELOPMENT OF LBB REQUIREMENTS 2.1. Evolution of LBB One of the many design criteria and requirements for the design of nuclear power units is referred to in Title 10, Part 50, Appendix A of the United States Code of Federal Regulations (10 CFR50 Appendix A) as General Design Criterion 4 (GDC-4). GDC-4 requires postulation of pipe breaks in nuclear power units and provision for appropriate protection against associated dynamic effects.However, in 1984 the US Nuclear Regulatory Commission (NRC) staff issued Generic Letter 84-04 (Reference

3) accepting that the double-ended guillotine break (DEGB) of the pressurized water reactor (PWR) primary loop piping was unlikely to occur, provided it could be demonstrated by deterministic fracture mechanics analyses that postulated small throughwall flaws in plant specific piping would be detected by the plant's leakage monitoring system long before the flaws could grow to unstable sizes. Leakage exceeding the limit specified in plant Technical Specifications requires operator action or plant shutdown.

The concept underlying such analyses is referred to as "leak-before-break" (LBB). A detailed discussion of limitations and acceptance criteria for LBB used by the NRC staff is provided in NUREG-1061, Volume 3 (Reference 4).2.2. Modifications to GDC-4 In July 1985, a proposed "limited scope" modification to GDC-4 was initiated by the NRC staff (Reference 5). A final "limited scope" rule was published on April 11, 1986 amending GDC-4 to permit the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures of primary coolant loop piping in PWRs (Reference 6). A proposed "broad-scope" modification to GDC-4 was initiated by the NRC staff in July 1986 (Reference 7). The final"broad-scope" rule was published on October 27, 1987 (Reference

8) which permitted the use of LBB analyses to all qualified high energy piping, i.e.pressure exceeding

-275 psi or temperature exceeding 200 F in nuclear power units.2.3. Development of New Standard Review Plan for LBB A new NRC Standard Review Plan (SRP) section, numbered 3.6.3 and entitled"Leak-Before-Break Evaluation Procedures" was published for public comment on August 28, 1987 in Volume 52, No. 167 of the Federal Register (Reference 9). As stated in References 8 and 9, the new SRP section will be used by the Page 4 FRAMATOME Doc. I.D. 51-5000709-00 NRC staff to review all submittals from licensees and applicants dealing with implementation of leak-before-break technology under the broad-scope amendment to GDC-4 of Appendix A, 10 CFR Part 50. The new SRP section 3.6.3 had its genesis in the USNRC Piping Review Committee report, NUREG-1061, Volume 3, dated November 1984, but has undergone substantial revision as a result of the efforts of the NRC staff, NRC contractors and consultants, the Committee to Review Generic Requirements (CRGR), the Advisory Committee on Reactor Safeguards (ACRS) and comments from the nuclear industry.

Since its issue for public comment, the new SRP 3.6.3 has not been revised in nearly 10 years. Based on the above, the new SRP 3.6.3 forms the basis for NRC staffs review of LBB submittals.

Page 5 TNAMATOM I Doe. I.D. 51-5000709-00

3. FATIGUE FLAW GROWTH ANALYSIS -TLAA ISSUE 1 The fatigue flaw growth analysis performed in the LBB analysis of the RCS primary piping has been identified as a TLAA issue. This section addresses this issue for the period of extended operation.

3.1. LBB Analysis of RCS Primary Piping As previously discussed, the generic LBB analysis report, reviewed and approved by the NRC, is contained in BAW-1847, Revision 1. This report is applicable to the B&W designed plants of all the GLRP participating utilities except for General Public Utilities (GPU) TMI-1 plant. The LBB report for TMI-1, also reviewed and approved by the NRC, is contained in BAW-1999 dated April 1987.The generic LBB analyses were performed during the 1984 to1985 time period and were therefore based on modified GDC-4 limited scope rule with the detailed analyses performed according to the limitations and acceptance criteria for LBB as given in NUREG 1061 Vol. 3. This is also true for the TMI-1 LBB report prepared in April of 1987. The limited scope rule permitted the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures of primary main coolant loop piping in PWRs.3.2. Acceptance Criteria for LBB Submittals The NRC staffs acceptance criteria for LBB submittals are contained in Section 5.0 of NUREG 1061 Volume 3. The fatigue flaw growth analysis requirement is given in item (d) of Section 5.2 of the NUREG. By this requirement, a surface flaw needs to be postulated at the location(s) of the piping system, identified for LBB analysis (highest stress coincident with the poorest material properties for base material, weldments and safe-ends).

At this location, a flaw size that would be permitted by the acceptance criteria of Section XI of the ASME Boiler and Pressure Vessel Code needs to be postulated.

A fatigue crack growth analysis for Code Class 1 piping then needs to be performed to demonstrate that the crack will not grow significantly during service. Additional details of this requirement is provided in Section 5.6 of NUREG 1061 Volume 3.In October of 1987, a final rule on the broadscope modification to General Design Criterion 4 of Appendix A, 10 CFR Part 50 was published.

The broadscope rule permits the use of analyses to exclude dynamic effects of pipe Page 6 F AMATOM E0 Doc. I.D. 51-5000709-00 ruptures in all high energy piping in all nuclear power units. The acceptance criteria section of the broadscope rule states that the Commission has"developed a new Standard Review Plan Section 3.6.3 which gives more details on how applicant and licensee submittels will be evaluated." The new SRP 3.6.3 was issued in August of 1987 for public comment. SRP 3.6.3 outlines current NRC staffs review procedures and acceptance criteria for LBB licensing applications.

3.3. Elimination of Requirements for Fatigue Crack Growth Analysis The new SRP section 3.6.3 had its genesis in the USNRC Piping Review Committee Report, NUREG-1061, Volume 3, dated November 1984. However, it has undergone substantial revision as a result of the efforts of the NRC staff, NRC contractors and consultants, the Gommittee to Review Generic Requirements (CRGR), and the Advisory Committee on Reactor Safeguards (ACRS). Additionally, nuclear industry comments on the proposed broadscope amendment to GDC-4 also had major impact on the new SRP.SRP 3.6.3 has not been updated since it was issued nearly ten years ago. The most notable change in the requirement, per SRP 3.6.3, is that fatigue crack growth analysis is not a requirement in the submittal for LBB analysis.Page 7 F RAMATOME Doec. I.D. 51-5000709-00 TI C N N LOSSES 4. THERMAL AGING -TLAA ISSUE 2 The susceptibility of the RCS primary piping to thermal aging was qualitatively addressed in the LBB analysis of the RCS primary piping. Thermal aging is identified as a TLAA issue and is addressed in this section for the period of extended operation.

4.1. RCS Primary Piping The RCS primary piping is primarily constructed from carbon s.'9el which has been clad with austenitic stainless steel or Alloy 82, 1.32. The safe ends at the suction and discharge ends of the RC pump casings, as illustrated in Figure 3-11 of Reference 1 are made of wrought stainless steel. Ic has been demonstrated in Section 3.1. of the NRC approved B&W Owners Group GLRP RCS piping report (Reference

10) that loss of fracture toughness due to thermal embrittlement is not an applicable aging effect on these components.

Thermal aging is therefore not a concern for the LBB analysis of the RCS primary piping (also referred to as RCS piping LBB analysis), since only the base metals, safe ends ani weldments of the RCS primary piping need to be considered in the evaluation for LBB (References 4, 9, 15).4.2. Cast Austenitic Stainless Steel (CASS) Components The RC pump casings are the only CASS components within the RCS primary piping of the B&W designed NSS. Test data obtained by Argonne National Laboratory (Reference 11), indicate that prolonged exposure of CASS to reactor coolant opErating temperatures can lead to thermal aging embrittlement.

The relevant aging effect on CASS components is the reduction in the fracture toughness of the material as a function of time. The magnitude of the reduction depends upon the casting method (statically or centrifugally cast), material chemistry (e.g. delta ferrite and molybdenum content) and the duration of exposure at coolant operating temperature.

The RC pump casings of ANO-1, ONS-2 and ONS-3 were fabricated from SA-351 Grade CF8M casting material.

SA-351 Grade CF8 casting material was used in the fabrication of the RC pump casings of ONS-1 and TMI-1. The materials of construction of the RC pump casings are reported in Table 2-1 of Reference

12. It is assumed that all of these pump casings are fabricated from statically cast materials.

This assumption is conservative since the fracture Page 8 TA ,E, Doc. I.D. 51-5000709-00 toughness of statically cast material is lower than the fracture toughness of centrifugally cast materials.

4.2.1. Justification in LBB Submittal The LBB submittal (which summarizes the RCS piping LBB analysis) qualitatively addressed thermal aging of the RC pump casings which are CASS components.

This assessment was included in the LBB submittal even though the RC pump casings did not require to be addressed for the LBB analysis of the RCS primary piping. However, the application of the LBB concept can be extended to components such as the RC pump casings.At the time of the preparation of the Generic LBB report, there was limited research done on thermal aging of CASS components.

Based on the information available at the time, a qualitative assessment was made as given in Section 3.3.4.3 of that report. The following is summarized from this assessment.

The reported values for aged cast stainless steel were considered to be encompassed by the values selected for the LBB evaluation as representative design values. Based on the above information and representative material properties in the RCS piping LBB aniysis, it was believed unnecessary to further consider the effects of thermal aging on material properties of RCS piping materials.

Finally, it ws stated that since these materials are widely used in similar elevated temper'ture a')lications, further information and service experience on the effects of thermai aging will become available.

4.2.2. Recent Development The recent developments on the issue of thermal aging of CASS components is summarized by S. Lee et al of the NRC and 0. Chopra of Argonne National Laboratory (ANL) in Reference

13. The information given below is excerpted from this reference.

In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials.

The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a database, both from data within ANL and from international sources, of about 85 compositions of cast stainless steel exposed to a temperature range of 550 -750 F for up to 58,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (6.5 years). From this database, ANL developed correlations for estimating the extent of thermal aging of cast stainless steel (References 11 and 14).Page 9 f FRAM ATOM E Ic N O LOGIIs Doc. I.D. 51-5000709-00 Based on the USNRC staffs review of the results of the ANL research program, discussed above, the UStIRC staff concl, ded that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (Reference 13). Thus the USNRC staff has accepted the use of SAW flaw evaluation procedure in IWB-3640 of Section Xl of the ASME Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation.

4.2.2.1. Fracture Toughness Estimation Procedures ANL developed fracture toughness estimation procedures for thermally aged cast stainless steel based on correlating experimental data (ANL's database) in a conservative manner. The estimation procedures are described in a flow chart with three distinct options, called "lower bound," "saturation," and "service time." The selection of different options would depend on the information available for the specific material and the special interest of the utility.The first and simplest option provides a lower bound estimate.

No information from the certified material test report (CMTR) of the CASS component is necessary for estimating the lower-bound fracture toughness.

However, if the ferrite content of the component is measured in the field using a ferrite gauge or is calculated based on the chemistry composition, this option provides a lower-bound estimate specifically for materials containing a similar range of ferrite.The other two options are more elaborate and require the use of information from the CMTRs. The second op" u,., provides an estimate for the "saturation" toughness, that is, the minimum fracture toughness remaining after thermal exposure for an infinitely long period of time. However, in some instances,"saturation" may occur at a time after the cast stainless steel component is no longer in service. Then, it may be beneficial to estimate the toughness of the component at some time prior to "saturation," and this is addressed by the third option for a specified time in service.4.2.3. Assessment of Fracture Toughness Properties of CASS Materials A comparison between the fracture toughness properties used in the RCS piping LBB analysis (Reference

1) and the lower-bound estimate of fracture toughness of CASS materials is addressed below.4.2.3.1. Fracture Toughness Properties for RCS Piping LBB Analysis Two sets of bounding fracture toughness properties were used in the RCS piping LBB analysis.

The two sets of data represent both the weld and base metals Page 10 FRAMATOME De. .D. 51-5000709-00 TZ C¢H NO LOG IIS used in the fabrication of the RCS piping in the B&WOG plants. The fracture toughness properties are given in Section 4.4 and summarized below.The lowest J-R curve for the weld metal was represented by a five parameter equation of the following form: JR = -6.51393(0.085

+ Aa)= + 3687.37Aa

+ 119794 where JR is material Jmo,0dgf in in-lb/in 2 and Aa is crack extension in inches. The J-R data for the base metal was represented by the following J-R equation.JR = -3.12872(0.06

+ Aa)-2 + 5172.63Aa

+ 1146.48 where JR and Aa are as defined above.4.2.3.2. Frarture Toughness Properties of CASS Materials The "lower bound" option of the ANL procedures, discussed in Section 4.2.2.1, provides a lower bound estimate for the thermally aged cast stainless steel as a J-R curve of the following form: Jd = C(Aa)°where Jd is the Deformation J in kJ/m 2 , Aa is crack extension in mm and the parameters C and n for lower-bound J-R curve are given in Table 1 of Reference 13 and in Section 3.1 of Reference

14. For LBB evaluation, the fracture toughness curve at normal operating temperature (550 IF) is applicable.

As previously discussed in Section 4.2, the RC pump casings of all 'he GLRP participating plants are fabricated using CF-8 or CF8M statically cast materials.

If the ferrite content for these pump casings are conservatively assumed to be greater than 15%. than the lower-botun

ý fracture toughness curve considered to envelop the RC pump casings of all the GLRP participating plants is given by Jd = ;87(Aa)0 3 1 where C = 167 and n = 0.31 are applicable for CF8M material at 550 OF with a ferrite content > 15%. As ,-nted in Section 3.1 of Reference 14, the cast stainless steel used in the U.S. nuclear industry generally contain less than 15%ferrite content and therefore the use of lower-bound fracture toughness properties may be very conservative.

Page I I IRAFAC T4 ,, 0oLO M,1, Doc. I.D. 51-5000709-00 4.2.3.3. Comparison of Fracture Toughness Data The fracture toughness curves for the lower bound base and weld metals used in the RCS piping LBB analysis and the fracture toughness curve for the lower-bound CASS material is illustrated in Figure 1. The fracture toughness curves were determined using the J-R equations given in Sections 4.2.3.1 and 4.2.3.2.The fracture toughness curve of the CASS material is clearly lower-bounding when compared against the fracture toughness curves used in the RCS piping LBB analysis.4.2.4. Approach for Evaluation of CASS Pump Casings The most limiting material and location used in the RCS piping LBB analysis was determined to be the base metal material of the straight section of the 28 inch cold ley pipe. Both the suction and discharye nozzles of the RC pump casings are attached to the 28 inch cold leg pipes and have similar geometry and loading applied to them as the limiting location used for the LBB analysis.

It was therefore concluded that the discharge and suction nozzles of the RC pump casings be evaluated using the LBB concept with lower-bound CASS fracture toughness properties given in Section 4.2.3.2. The results of this evaluation is given in Section 4.2.5 below.4.2.5. Evaluation of CASS Pump Casing Nozzles The bounding 10 gpm leakage crack sizes for the suction and discharge nozzle were determined in Reference

16. The leakage crack length (twice the leakage flaw size) for the suction nozzle was determined to be 8.62 inches and the leakage crack length for the discharge nozzle was determined to be 8.86 inches.The flaw stability analysis for the CASS pump casing nozzles (suction and discharge) were determined in Reference

17. To account for thermal aging of the nozzles during the period of extended plant operation, the analysis conservatively utilized the lower-bound J-R curve for aged CF8M material from Reference

13. The most limiting nozzle was determined to be the discharge nozzle. For this nozzle, the maximum applied J, for the 10 gpm leakage flaw size, was determined to be 0.510 kips/in. The margin on flaw size was determined to be 2.4 which is greater than the required margin of 2 per S.R.P.3.6.3.Page 12 fFRAMATOM E TI[ C H N 0100 t 5 Doc. I.D. 51-5000709-00 Figure 1. Comparison of J-R Data I C 0 C 900 800 700 600 500 400 300 200 100~CF8Mdf>15%

SLBB-Base Metal 0 0 2 4 6 8 10 12 14 16 Crack Extension

(,nm)18 20 Page 13 W MFRAMATOM E Doc. I.D. 51-5000709-00

5.

SUMMARY

There are only two TLAA issues involved in the LBB analysis of the RCS primary piping whose results are summarized in LBB reports of References 1 and 2.The first TLAA issue is the fatigue flaw growth analysis.

At the time the LBB analyses were performed, fatigue flaw growth analysis was required per Reference

4. By this requirement, a flaw had to be postulated at the selected LBB location of the piping to demonstrate that the postulated flaws will not grow significantly during service. However, subsequently, the NRC issued a new SRP 3.6.3 (Reference

9) which forms the basis for NRC staffs review of LBB submittals.

The new SRP 3.6.3 does not contain any requirement to perform fatigue flaw growth analysis.

It is therefore concluded that, fatigue flaw growth analysis is not required for justifi'..ation of LBB of the RCS primary piping for the period of extended operation.

The second TLAA issue is thermal aging of the RCS primary piping. The RCS primary piping is pr~inarily constructed from carbon steel which has been clad with austenitic stainless steel or Alloy 82/182. It has already been demonstraied in the NRC approved B&W Owners Group GLRP RCS piping report (Reference

10) that loss of fracture toughness due to thermal embrittlement is not an applicable aging effect on these components.

However, the RC pump casings which are fabricated from statically cast CF-8 or CF8M CASS materials, for the GLRP participating plants, are susceptible to thermal embrittlement after prolonged exposure at reactor operating temperatures.

The flaw stability analysis in support of LBB has been demonstrated for the RCP nozzles of the GLRP participating utilities (Reference 17). Based on this analysis it was determined that the RCP nozzles, with consideration of thermal aging, meets all the safety margin requirements of S.R.P. 3.6.3. It is therefore demonstrated that thermal aging of CASS RCP casing nozzles of ANO-1, ONS-1, ONS-2, ONS-3 and TMI-1 is acceptable for the period of extended plant operation in accordance with 54.21 (c)(1)(ii).

In summary, the generic LBB analyses for the B&W operating plants reported in BAW-1 847, Revision 1, and BAW-1 999 remain valid for the period of extended operation with the exception of the assessment of reduction of fracture toughness by thermal aging of cast austenitic stainless steel RCP nozzles.Reduction of fracture toughness of the RCP nozzles was determined to be acceptable for the period of extended operation through the flaw stability analysis described above.Page 14 T AMATroM N Doc. I.D. 51-5000709-00

6. REFERENCES

1. "The B&W Owners Group Leak-Before-Break Evakiation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS," BAW-1847, Revision 1. September 1985.2. "TMI-1 Nuclear Power Plant Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping," BAW-1999, April 1987.3. Generic Letter 84-04. Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops. U.S. NRC Commission, Washington DC, February 1984.4. NUREG-1061, Vol. 3. Evaluation of Potential for Pipe Breaks, Report of the US Nuclear Regulatory Commission, Washington DC, November 1984.5. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures.

Federal Register, Vol. 50, No. 126, Published by the Office of Federal Register, Washington DC, July 1, 1985, pp. 27006-9.6. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures.

Federal Register, Vol. 51, No. 70, Published by the Offict-: of Federal Register, Washington DC, April 11, 1986, pp. 12502-5.7. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures.

Federal Register, Vol. 51, No. 141, Published by the Office of Federal Register, Washington DC, July 23, 1986, pp. 26393-9.8. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Pos!ulated Pipe Ruptures.

Federal Register, Vol. 52, No. 207, Published by the Office of Federal Register, Washington DC, October 27, 1987, pp. 41288-95.9. Standard Review Plan 3.6.3. Federal Register Volume 52, Number 167, Published by the Office of Federal Register, Washington DC, August 28, 1987, pp. 32626-33.Page 15 T RAMATOM I Doc. I.D. 51-5000709-00

10. "Demonstration of the Management of Aging Effects for the Reactor Coolant System Piping," BAW-2243A, June 1996.11. 0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, U.S. Nuclear Regulatory Commission, Washington DC, May 1994.12. FTI Document 77-1258459-00, "Demonstration of the Management of Aging Effects for the Reactor Coolant Pump," December 1996.13. S. Lee, P. T. Kuo, K. Wichman of U.S. Nuclear Regulatcry Commission and 0. Chopra of Argonne National Laboratory, "Flaw Evaluation of Thermally Aged Cast Stainless Steel in Light-Water Reactor Applications," International Journal of Pressure Vessel and Piping, to be published.

14. 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," NUREGICR-4513, Revision 1, ANL-93/22 U.S. Nuclear Regulatory Commission, Washington DC, August 1994.15. USNRC Safety Evaluation Report, "Safety Evaluation of B&W Owners Group Reports Dealing With Elimination of Postulated Pipe Breaks in Primary Main Loops," by Dennis M Crutchfield, Assistant Director for Technical Support Division of PWR Licensing-B of Office of Nuclear Reactor Regulation to L. C. Oakes, Chairman of B&W Owners Group Leak-Before-Break Task Force, dated December 12, 1985.16. FTI Document 32-5000940-00, "RCP Nozzle Loads & Leak-Rate Analysis".

17. FTI Document 32-5000994-00, "FM Analysis for LBB of RCP Nozzles".Page 16