RS-14-168, Responses to NRC Requests for Additional Information, Set 26, Dated May 21, 2014, Related to License Renewal Application

Responses to NRC Requests for Additional Information, Set 26, Dated May 21, 2014, Related to License Renewal Application

ML14168A020
Person / Time
Site:	Byron, Braidwood
Issue date:	06/16/2014
From:	Gallagher M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-168
Download: ML14168A020 (16)

Text

Michael P. Gallagher Vice President, License Renewal Exelon Nuclear Exelon Generation 200 Exelon Way Kennett Square. PA 19348 610 765 5958 Office 610 765 5956 Fax www.exeloncorp.com michaelp.gallagher@exeloncorp.com 10 CFR 50 10CFR51 10 CFR 54 RS-14-168 June 16, 2014 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Responses to NRC Requests for Additional Information, Set 26, dated May 21, 2014, related to the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application

References:

1. Letter from Michael P. Gallagher, Exelon Generation Company LLC (Exelon) to NRC Document Control Desk, dated May 29, 2013, "Application for Renewed Operating Licenses"

2. Letter from Lindsay R. Robinson, US NRC to Michael P. Gallagher, Exelon, dated May 21, 2014, "Request for Additional Information for the Review of the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Application, Set 26 (TAC NOS. MF1879, MF1880, MF1881, and MF1882)"

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted the License Renewal Application (LRA) for the Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (BBS). In Reference 2, the NRC requested additional information to support staff review of the LRA.

Enclosure A contains the responses to these requests for additional information.

Enclosure B contains updates to sections of the LRA affected by the response.

There are no new or revised regulatory commitments contained in this letter.

June 16, 2014 U.S. Nuclear Regulatory Commission Page 2 If you have any questions, please contact Mr. Al Fulvio, Manager, Exelon License Renewal, at 610-765-5936.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on Respectfully,

Enclosures:

A. Responses to Requests for Additional Information B. Updates to affected LRA sections cc: Regional Administrator- NRC Region Ill NRC Project Manager (Safety Review), NRR-DLR NRC Project Manager (Environmental Review), NRR-DLR NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station NRC Project Manager, NRR-DORL-Braidwood and Byron Stations Illinois Emergency Management Agency - Division of Nuclear Safety

RS-14-168 Enclosure A Page 1 of 9 Enclosure A Byron and Braidwood Stations (BBS), Units 1 and 2 License Renewal Application Responses to Requests for Additional Information RAI 4.7.4-1a RAI 3.1.1.81-1a

RS-14-168 Enclosure A Page 2 of 9 RAI 4.7.4-1a Applicability:

Byron Station (Byron) and Braidwood Station (Braidwood), Units1 and 2

Background:

By letter dated March 28, 2014, the applicant responded to request for additional information (RAI) 4.7.4-1 which addresses the applicants fracture mechanics analysis for the flaws detected in the tube side inlet and outlet nozzles of residual heat removal (RHR) heat exchangers. As part of the response, the applicant provided information regarding when it performed the most recent volumetric examinations on each nozzle of the Byron and Braidwood RHR heat exchangers. The applicant also indicated that the most recent volumetric examinations for the RHR heat exchanger nozzles were those performed in 1994 on Braidwood, Unit 2, RHR heat exchanger nozzles.

During its review of the applicants response and related information, the staff noted NRC letter dated February 29, 1996 (ADAMS Accession No. 9603060023), which encloses the staffs safety evaluation regarding the Byron and Braidwood request for relief (Nos. NR-18 and NR-23) from the volumetric examinations of the RHR heat exchanger nozzles for the first 10-year inservice inspection interval. This safety evaluation discusses the previous inspection requirements which are specified in the staffs safety evaluation, dated February 3, 1995 (ADAMS Accession No. 9502130021), regarding the flaws detected in these nozzles and the applicants fracture mechanics analysis for the flaws subject to the evaluation of American Society of Mechanical Engineers (ASME)Section XI, IWB-3600.

The February 29, 1996, safety evaluation further states that instead of the previous requirements specified in the February 3, 1995, safety evaluation, the licensee is required to perform ultrasonic testing (UT) examinations on a sample of RHR nozzle-to-vessel welds (one nozzle per unit) during the next inspection interval (i.e., the second interval) to provide additional assurance that these flaws have not grown and that no new service induced indication has developed.

In its review of the applicants RAI response, the staff also noted that the applicants letter dated July 25, 2007 (ADAMS Accession No. ML072060413), describes a relief request regarding the Braidwood, Units 1 and 2, RHR heat exchanger nozzle examinations for the second 10-year inspection interval. The staff further noted that even though this 2007 relief request was withdrawn by the applicants letter dated January 23, 2008 (ADAMS Accession No.

ML080240324), the July 25, 2007, letter indicates that UT examinations were performed in September 1998, on a nozzle (1RHR-01-1RHXN1, A HX) of Braidwood, Unit 1, RHR heat exchangers to fulfill the requirements specified in the NRC safety evaluation dated February 22, 1996, to volumetrically inspect one nozzle per unit during the second inservice inspection interval. The applicants 2007 letter also states that no appreciable flaw growth was noted from the 1998 examinations on the examined nozzle of Braidwood, Unit 1.

RS-14-168 Enclosure A Page 3 of 9 Issue:

It is unclear to the staff why the applicants response to RAI 4.7.4-1 does not discuss the UT examination results for the Braidwood, Unit 1, RHR heat exchanger nozzle which were obtained in September 1998 as described in the applicants letter dated July 25, 2007. It is also unclear to the staff why the applicants response does not address any results of the UT examinations, which are associated with the applicants fracture mechanics analysis and are required for the RHR heat exchanger nozzles (i.e., a nozzle per unit) as specified in the staffs safety evaluation dated February 29, 1996.

Furthermore, it is unclear to the staff whether the previous examinations of the nozzles, including the 1998 examinations, indicate that any of these flaws grew. In addition, the staff needs clarification on whether the existing flaws are embedded inside the RHR heat exchanger nozzles without exposure to the reactor coolant in order to confirm the absence of environmental effects on flaw growth. The staff also noted that the applicants response did not provide the length of the bounding flaw with a depth of 0.300 inches.

Request:

1. Clarify why the applicants response to RAI 4.7.4-1 does not discuss the UT examination results of the Braidwood, Unit 1, RHR heat exchanger nozzle which were obtained in September 1998 as described in the applicants letter dated July 25, 2007.

2. Clarify why the applicants response does not address results of the UT examinations, which are associated with the applicants fracture mechanics analysis and are required for the RHR heat exchanger nozzles (i.e., a nozzle per unit) as specified in the staffs safety evaluation dated February 29, 1996.

If all of these UT examinations have not been completed, justify why the applicant does not identify the UT examinations as part of the 10 CFR Part 54.21(c)(1)(iii) aging management basis associated with the applicants fracture mechanics analysis.

3. Provide additional information to confirm whether the previous examinations of the nozzles, including the 1998 examinations, indicate that any of these flaws grew. As part of the response, define no appreciable flaw growth which was mentioned in the Byron and Braidwood letter dated July 25, 2007.

4. Clarify whether the existing flaws are embedded inside the RHR heat exchanger nozzles without exposure to the reactor coolant in order to confirm the absence of environmental effects on flaw growth. In addition, describe the length of the bounding flaw in comparison with the inner diameter of the nozzle.

RS-14-168 Enclosure A Page 4 of 9 Exelon Response:

1. RAI 4.7.4-1 Request 3 was interpreted as a request to provide the flaw size, orientation, and locations based the most recent (e.g. last performed) examinations for the period between 1991 and 1994 supporting the February 3, 1995 staff safety evaluation (SE). The results of these examinations were dispositioned using the acceptance criteria and flaw growth rates developed in the fracture mechanics analysis.

The fracture mechanics analysis, which was performed in accordance with ASME Section XI, paragraph IWB-3640, dispositioned flaws found between 1991 through 1994 in the Byron and Braidwood Unit 1 and 2 Residual Heat Removal (RHR) heat exchanger tube side inlet and outlet nozzle welds. During this time period, all RHR tube side inlet and outlet nozzle welds were ultrasonically examined at least once, and a number of nozzles were ultrasonically examined multiple times in multiple refueling outages. For example, the Braidwood Unit 2 A RHR heat exchanger inlet nozzle was ultrasonically examined on four (4) different occasions during the period from 1991 through 1994. The results of the fracture mechanics analysis were used to disposition the ultrasonic examination data collected from 1991 through 1994 in accordance with ASME Section XI, paragraph IWB-3640. The fracture mechanics analysis concluded that crack growth for the complete range of crack sizes would be inconsequential over the life of the plant (a total growth of less than 0.001 inches in 40 years). The fracture mechanics analysis included no further requirements or recommendations to perform follow-up ultrasonic examinations as validation of the conclusions.

The February 3, 1995 staff SE evaluated the submitted fracture mechanics analysis, the ultrasonic examination data for the period from 1991 through 1994, and other supporting information. The February 3, 1995 SE concluded that the methodology, including the use of the rules in ASME IWB-3640 for flaw evaluation, was appropriate.

Given the above, RAI 4.7.4-1 Request 3 was interpreted as a request to provide the flaw size, orientation, and locations based on the most recent inspections, between 1991 and 1994, that were dispositioned by the fracture mechanics analysis and evaluated in the February 3, 1995 SE. Request 3 to RAI 4.7.4-1 was not interpreted as a request for flaw size, orientation, and locations resulting from ultrasonic examinations after 1994. Therefore, the 1998 ultrasonic examination results were not included in the response.

2. As discussed above, the response to RAI 4.7.4-1 Request 3 was based on the most recent ultrasonic examination data, collected during the period between 1991 and 1994, that was dispositioned by the fracture mechanics analysis. Additional ultrasonic examinations after 1994 have been performed on Byron and Braidwood RHR heat exchangers nozzles in accordance with requirements of the February 29, 1996 SE referenced in the Background section of this RAI. This SE required ultrasonic examination of one RHR heat exchanger nozzle for each Byron and Braidwood Unit in the second 10-year Inservice Inspection (ISI) interval. The following table documents ultrasonic examinations performed in the second 10-year ISI interval (second interval) on the Byron and Braidwood Stations RHR heat exchanger nozzles in accordance with requirements of the February 29, 1996 SE and corresponding first 10-year ISI interval (first interval) examinations of the same nozzles.

RS-14-168 Enclosure A Page 5 of 9 Second Interval Ultrasonic Examination Required by the February 29, 1996 SE and Corresponding First Interval Examinations Unit Heat Inspection ISI Conclusions Exchanger/ Date Interval Nozzle Byron Unit 1 1RH02AA Spring 1993 First No observed flaw growth Inlet from first to second Byron Unit 1 1RH02AA Fall 1997 Second interval Inlet Byron Unit 2 2RH02AA Spring 1995 First No observed flaw growth Outlet from first to second Byron Unit 2 2RH02AA Spring 2001 Second interval Outlet Braidwood 1RH02AA Fall 1992 First No observed flaw growth Unit 1 Inlet from first to second Braidwood 1RH02AA Fall 1998 Second interval Unit 1 Inlet Ultrasonic examination of one Braidwood Unit 2 RHR heat exchanger nozzle was planned for the spring 2008 (A2R13) refueling outage, which was the last refueling outage in the Braidwood Unit 2 second interval. Because ultrasonic examination of these nozzles requires extensive labor resources, radiation exposure to the examiners, and significant cost without a commensurate increase in quality or public safety, a relief request was summited to the NRC in the letter dated July 25, 2007, which is referenced in the Background section of this RAI. The request asked the NRC for relief from the ASME Section XI requirement to perform ultrasonic examination of the Braidwood Station RHR heat exchanger nozzle per the alternative visual examination requirements in ASME Code Case N-706-1, which in July of 2007 had not yet been endorsed by the NRC. ASME Code Case N-706-1 provides relief from the requirement to perform ultrasonic examination of welds on PWR stainless steel regenerative and residual heat exchangers, provided the welds have been volumetrically examined at least once. Four (4) months later, in December 19, 2007, the NRC endorsed ASME Code Case N-706-1. Therefore, since all Braidwood Unit 2 RHR heat exchanger nozzles had been ultrasonically examined at least once during the period between 1991 and 1994, satisfying Code Case N-706-1 requirements, ultrasonic examination of the one Braidwood Unit 2 RHR heat exchanger nozzle in the second interval was no longer required in the spring 2008 refueling outage. On January 23, 2008 Exelon submitted a follow-up letter to the NRC withdrawing the relief request submitted in July of 2007. By letter dated January 31, 2008 the NRC acknowledged the withdrawal of the relief request.

Referring to the above table, flaws that were characterized by ultrasonic examinations in the second interval were individually compared by level three (3) NDE inspectors to the same flaws that were ultrasonically examined in the first interval. These comparisons concluded that there was no observed growth from the first interval to the second interval. Also, the second interval examinations found no new flaws. These ultrasonic examination results do not challenge the conclusions of the fracture mechanics analysis, which assumed 200 plant heatups and 200 plant cooldowns, and concluded that the RHR nozzle weld flaw growth

RS-14-168 Enclosure A Page 6 of 9 would be inconsequential (a total growth of less than 0.001 inches) over the life of the plant.

Since, the fracture mechanics analysis and ASME Code Case N-706-A, Table 1, note (2) require no additional follow-up ultrasonic examinations, the TLAA described in LRA section 4.7.4 does not depend on additional ultrasonic examinations. Therefore, additional UT examinations are not part of the 10 CFR Part 54.21(c)(1)(iii) disposition. LRA Section 4.7.4 documents that the 60-year projections of heatup and cooldown transients provided in LRA Section 4.3.1 have been demonstrated to be lower than the applicable transient limits. The Fatigue Monitoring Program will be used to ensure these limits are not exceeded.

3. As discussed above in response to Request 2, Byron and Braidwood Stations have performed ultrasonic examinations of some of the RHR heat exchanger nozzles in the second interval. All flaws examined in the second interval met the acceptance criteria established in the fracture mechanics analysis. The second interval ultrasonic examinations utilized the same ultrasonic techniques and similar equipment as the ultrasonic examinations performed in the first interval. Individual flaws characterized in the second interval were compared, by level three (3) NDE inspectors, to the same individual flaws characterized in the first interval. These comparisons concluded that there was no observed flaw growth greater than the repeatability variances of the examination equipment and techniques. The term no appreciable flaw growth in the July 25, 2007 letter was intended to explain that any dimensional differences between the flaw examination results in the second interval and the examination results of those same flaws in the first interval were small and within these repeatability variances. The repeatability variances include factors such as slight variations in the transducer placement angles, orientation, and surface contact during the scanning process, and slight differences in the scanners, search units, and gain levels.

4. The 1991 through 1994 first interval inspections reported in Exelons response to RAI 4.7.4-1 found that all indications are subsurface flaws, not opened to the internal and external surface of the nozzle, and therefore not exposed to reactor coolant. These flaws were determined to be fabrication flaws. The examinations performed in the second interval found that none of the flaws had grown and the flaws remain subsurface. In addition, periodic liquid penetrant examinations of all RHR heat exchanger nozzles have found no external flaws. As documented in the February 29, 1996 safety evaluation discussed in the Background section of this RAI, the fracture mechanics analysis concluded that the outside surfaces of the nozzle welds are in tension while the inside surfaces of the nozzle welds are in compression. Consequently, any service induced flaws would be expected to initiate on the outside surface of the nozzle welds.

As described in Exelons response to RAI 4.7.4-1, flaw number three (3) on the Braidwood Unit 2 2B RHR Heat Exchanger inlet nozzle bounds all flaws that were dispositioned as acceptable for continued service in accordance with ASME Section XI, paragraph IWB-3640. Review of the second interval ultrasonic examination results shows that this flaw remains the bounding flaw. This flaw has an as found crack depth of 0.300 inches in a portion of the nozzle with a wall thickness of 0.526 inches. Applying the fatigue flaw crack depth growth of 0.001 inches, after an additional 200 cycles, results in a projected crack depth of 0.301 inches and the percentage of flaw depth to nozzle wall thickness would be 57.2%, which meets the acceptance criterion of 60%. The length of this flaw was 0.8 inches. The inner diameter of this nozzle is 13.075 inches. This value is based on 14-inch

RS-14-168 Enclosure A Page 7 of 9 schedule 40 pipe, which is machined down to 13.875 inches, with an actual ultrasonically-measured wall thickness of 0.4 inches.

RS-14-168 Enclosure A Page 8 of 9 RAI 3.1.1.81-1a Applicability:

Byron and Braidwood

Background:

By letter dated March 28, 2014, the applicant responded to RAI 3.1.1.81-1, which addressed LRA item 3.1.1-81. This LRA item states that the Water Chemistry program and One-Time Inspection program are used to manage cracking due to stress-corrosion cracking (SCC) for stainless steel pressurizer spray heads exposed to reactor coolant. LRA item 3.1.1-81 corresponds to NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants Table 3.1-1, item 81, which contains GALL Report item IV.C2.RP-41.

Issue:

During its review of the applicants response and related information, the staff noted that LRA Table 3.1.2-1 for the reactor coolant system does not include a specific aging management review (AMR) line item for GALL Report item IV.C2.RP-41, which manages cracking due to SCC of stainless steel pressurizer spray heads in accordance with LRA item 3.1.1-81.

Therefore, the staff cannot determine how the applicant will manage cracking due to SCC of pressurizer spray heads.

Request:

Clarify why LRA Table 3.1.2-1 does not include a specific AMR line item which manages cracking due to SCC of stainless steel pressurizer spray heads using LRA item 3.1.1-81.

Alternatively, revise the LRA to identify an AMR line item which is associated with LRA item 3.1.1-81 to manage cracking due to SCC for these components.

Exelon Response:

Upon further review of the current licensing basis for the crediting of the Chemical & Volume Control System (CVCS) auxiliary spray, aging management review (AMR) line items associated with the Reactor Coolant System pressurizer spray head are added to the LRA. The aging effect of cracking will be managed by the Water Chemistry and One-Time Inspection aging management programs. The aging effect of loss of material will be managed by the Water Chemistry aging management program.

The function of the pressurizer spray head is to provide a method of pressure control by condensing the pressurizer steam bubble. The pressurizer spray head is fabricated from ASTM A-269 CF8M cast austenitic stainless steel (CASS). Aging management of loss of fracture toughness due to thermal aging embrittlement of CASS is not required since the pressurizer spray head is not part of the Class 1 reactor coolant pressure boundary, is not a pressure-retaining component, and is not a structural component. The pressurizer spray head does not perform an intended function at low temperatures. The stresses on the pressurizer spray head at low temperatures are negligible and no mechanism exists to apply excessive

RS-14-168 Enclosure A Page 9 of 9 loading, therefore, the pressurizer spray head is not subject to loads that would result in a fracture (non-ductile failure) at low temperatures. NRC Letter from C.I. Grimes, NRC, to Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components, dated May 19, 2000 (Accession Number ML003717179), supports this conclusion. The NRC letter states, If the loading is compressive or low enough to preclude fracture of the component, then the component would not require supplemental inspection.

LRA Tables 2.3.1-1, 3.1.1, and 3.1.2-1 are revised as shown in Enclosure B.

RS-14-168 Enclosure B Page 1 of 5 Enclosure B Byron and Braidwood Stations, Units 1 and 2 License Renewal Application Updates to Affected LRA Sections Resulting from the responses to the following RAI:

RAI 3.1.1.81-1a Note: To facilitate understanding, portions of the original LRA have been repeated in this Enclosure, with revisions indicated. Existing LRA text is shown in normal font. Changes are highlighted with bolded italics for inserted text.

RS-14-168 Enclosure B Page 2 of 5 As a result of the response to RAI 3.1.1.81-1a, LRA Table 2.3.1-1, page 2.3-11 is revised as shown below. Revisions are indicated with bolded italics for inserted text:

Component Type Intended Function Pressurizer Spray Head Spray Pressurizer surge and steam space Pressure Boundary nozzles, and welds (Class 1)

Pump Casing (Reactor Coolant Pump Pressure Boundary Class 1)

Pump Casing (Reactor Coolant Pump Leakage Boundary Motor Oil Lift Pump)

Reactor Coolant Pressure Boundary Pressure Boundary Components (Hot Leg, Intermediate Leg, Cold Leg, and Class 1 Piping >4" NPS)

Restricting Orifice Pressure Boundary Throttle Restricting Orifice (Class 1) Pressure Boundary Throttle Rupture Disks Leakage Boundary Strainer Body (Reactor Coolant Pump Leakage Boundary Motor Oil Lift Component)

Tanks (Pressurizer Relief Tank) Leakage Boundary Tanks (Reactor Coolant Pump Motor Leakage Boundary Upper and Lower Oil Reservoirs, integral to motor)

Valve Body Leakage Boundary Pressure Boundary Structural Support Valve Body (Class 1) Pressure Boundary The aging management review results for these components are provided in:

Table 3.1.2-1 Reactor Coolant System Summary of Aging Management Evaluation

RS-14-168 Enclosure B Page 3 of 5 As a result of the response to RAI 3.1.1.81-1a, LRA Table 3.1.1, page 3.1-45 is revised as shown below. Revisions are indicated with bolded italics for inserted text:

Table 3.1.1 Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Programs Evaluation Recommended 3.1.1-81 Stainless steel Cracking due to Chapter XI.M2, Water No Consistent with NUREG-1801. The One-pressurizer spray head stress corrosion Chemistry, and Chapter Time Inspection (B.2.1.20) program and exposed to reactor cracking XI.M32, One-Time Water Chemistry (B.2.1.2) program will be coolant Inspection used to manage cracking of the stainless steel pressurizer spray head and heat exchanger components exposed to reactor coolant in the Reactor Coolant System.

RS-14-168 Enclosure B Page 4 of 5 As a result of the response to RAI 3.1.1.81-1a, LRA Table 3.1.2-1, page 3.1-60 is revised as shown below. Revisions are indicated with bolded italics for inserted text:

Table 3.1.2-1 Reactor Coolant System (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Item Notes Type Function Requiring Programs Item Management Pressurizer Spray Spray Cast Austenitic Reactor Coolant Cracking One-Time Inspection IV.C2.RP-41 3.1.1-81 A,4 Head Stainless Steel (External) (B.2.1.20)

(CASS)

Water Chemistry (B.2.1.2) IV.C2.RP-41 3.1.1-81 A,4 Loss of Material Water Chemistry (B.2.1.2) IV.C2.RP-23 3.1.1-88 A Cast Austenitic Reactor Coolant Cracking One-Time Inspection IV.C2.RP-41 3.1.1-81 A,4 Stainless Steel (Internal) (B.2.1.20)

(CASS) Water Chemistry (B.2.1.2) IV.C2.RP-41 3.1.1-81 A,4 Loss of Material Water Chemistry (B.2.1.2) IV.C2.RP-23 3.1.1-88 A

RS-14-168 Enclosure B Page 5 of 5 As a result of the response to RAI 3.1.1.81-1a, LRA Table 3.1.2-1, page 3.1-69 is revised as shown below. Revisions are indicated with bolded italics for inserted text:

Table 3.1.2-1 Reactor Coolant System (Continued)

Plant Specific Notes:

4. The Reactor Coolant System pressurizer spray head is non-pressure retaining, non-structural, does not perform an intended function at low temperatures, and is not subjected to loads that would result in a fracture (non-ductile failure) at low temperatures. No additional inspection activities are required to manage the loss of fracture toughness due to thermal aging embrittlement.