View source for TMI-09-142, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations - Relief Requests RR-09-01 and RR-09-02

{{Adams
| number = ML093020523
| issue date = 10/29/2009
| title = Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations - Relief Requests RR-09-01 and RR-09-02
| author name = Cowan P
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000289
| license number = DPR-050
| contact person = 
| case reference number = TAC MC9768, TMI-09-142
| document type = Letter type:TMI
| page count = 19
| project = TAC:MC9768
| stage = Request
}}

=Text=
{{#Wiki_filter:10 CFR 50.55a TMI-09-142 October 29, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

==Subject:==
Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations - Relief Requests RR-09-01 and RR-09-02

==References:==
: 1) Pressurized Water Reactor Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated June 2008
: 2) Letter from H. K. Nieh (NRC) to G. Bischoff (PWROG), "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval' (TAC No. MC9768)," dated May 8,2008 In Pressurized Water Reactor Owners Group (PWROG) Topical Report WCAP-16168-NP-A, Revision 2 (Reference 1), the PWROG provided the technical and regulatory basis for decreasing the frequency of inspections by extending the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI inservice inspection interval from the current 10 years to 20 years for Examination Categories B-A and B-D reactor pressure vessel (RPV) welds. The U.S. Nuclear Regulatory Commission (NRC) approved the topical report by letter dated May 8,2008 (Reference 2). To implement the change presented in Reference 1, Exelon Generation Company, LLC (EGC) hereby submits Attachment 1 (RR 01) in accordance with the Safety Evaluation (Reference 2), to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. The plant specific information identified by Reference 1 to support this proposed alternative for Three Mile Island (TMI), Unit 1 is provided in Attachment 1. The next inspection of Examination Categories B-A and B-D RPV welds are scheduled to be performed in the 2015 refueling outage.

Request to Extend the lSI Interval for Reactor Vessel Weld Examinations Relief Requests RR-09-01 and RR-09-02 October 29, 2009 Page 2 Further, pursuant to 10 CFR 50.55a(a)(3)(ii), TMI, Unit 1 is also requesting approval to delay compliance with the specified requirements of ASME B&PV Code, Section XI, Table IWB-2500-
: 1. Relief is requested to defer visual examinations under Table IWB-2500-1, Examination Categories B-N-2 and B-N-3, Item Nos. B13.50 and B13.70 for the extended inspection interval from 10 to 20 years as described in Attachment 2 (RR-09-02). Deferring the subject visual examinations associated with the reactor vessel core support structure removal to the extended inspection interval allows those examinations to be performed at the same time (2015 refueling outage) as the Examination Categories B-A and B-D RPV welds described in Attachment 1.
In the Safety Evaluation (SE) dated May 8,2008 for Topical Report WCAP-16168-NP, Revision 2 (Reference 2), the NRC required licensees to amend their facility operating license to provide the NRC with the information and analyses requested in Section (e) of the final rule for paragraph 10 CFR 50.61 (a) or the proposed rule published in the Federal Register on October 3, 2007 (72 FR 56275) within one year, following completion of each ASME Code, Section XI, Category B-A and B-D weld inspection. EGC understands this requirement is no longer necessary because the NRC will grant inservice inspection interval extensions on an interval-by-interval basis. Based on this NRC position, the NRC granted in 74 FR 29247 a request by Waterford Steam Electric Station, Unit 3 to withdraw its license amendment request required by this SE.
EGC requests approval of these requests by October 29,2010.
There are no regulatory commitments in this letter.
If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.
Respectfully, Pamela B. Cowan Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Relief Request RR-09-01
: 2) Relief Request RR-09-02 cc:     S. J. Collins, Regional Administrator, Region I, USNRC D. M. Kern, USNRC Senior Resident Inspector, TMI P. Bamford, Project Manager [TMI] USNRC

Attachment 1 Relief Request RR-09-01

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 1 of 11)
RR-09-01 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
              -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Component's) Affected These examination categories and item numbers are from Table IWB-2500-1 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code, Section XI:
Examination Category         Item No.       Component                      Description B-A             B1.11          RCT0001 RV0012                 Circumferential Shell Welds RCT0001 RV0013 RCT0001 RV0014 RCT0001 RV0015 B-A              B1.12          RCT0001 RV0018L                Longitudinal Shell Welds RCT0001 RV0019L RCT0001 RV0020L RCT0001 RV0021 L B-A              B1.21          RCT0001 RV0016                 Circumferential Head Welds B-A              B1.30          RCT0001 RV0011                 Shell-to-Flange Weld B-A              B1.51          RCT0001 A12052                 Beltline Region Repair Weld B-D              B3.90          RCT0001 RV0001 N               Nozzle-to-Vessel Welds RCT0001 RV0002N*
RCT0001 RV0003N RCT0001 RV0004N RCT0001 RV0005N*
RCT0001 RV0006N RCT0001 RV0007N RCT0001 RV0008N B-D              B3.100         RCT0001 RV0001N                Nozzle Inner Radius Areas RCT0001 RV0002N*
RCT0001 RV0003N RCT0001 RV0004N RCT0001 RV0005N*
RCT0001 RV0006N RCT0001 RV0007N RCT0001 RV0008N (Throughout this request the above examination categories are referred to as "the sUbject examinations" and the ASME B&PV Code, Section XI, is referred to as "the Code.")
* Note that the third 10-year Inservice Inspection (lSI) interval examinations for Identification RCT0001 RV0002N and RCT0001 RV0005N were completed in the 2001

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 2 of 11) refueling outage. The remaining examinations during the 2001 refueling outage were credited as second 10-year lSI interval examinations. This is allowed per IWA-2430(d)(2).

===2. Applicable Code Edition and Addenda===
The applicable code edition and addenda for Three Mile Island (TMI), Unit 1 is the ASME B&PV Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1995 Edition with Addenda through 1996 (Reference 1) for the third 10-year lSI interval. The applicable ASME B&PV Code, Section XI Edition and Addenda for subsequent intervals will be as required by 10 CFR 50.55a.

===3. Applicable Code Requirement===
IWB-2412, Inspection Program B, requires volumetric examination of essentially 100%
of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each 10-year interval. The TMI, Unit 1 third 10-year lSI interval is scheduled to end April 19, 2011; however, TMI will be applying Section XI, IWA-2430(d)(1) to extend the current interval through the fall 2011 refueling outage (T1 R19).

===4. Reason for Request===
An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval.
Extension of the inspection interval for Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
: 5. Proposed Alternative and Basis for Use Exelon Generation Company, LLC (EGC) proposes to defer the ASME B&PV Code, Section XI required volumetric examination of the TMI, Unit 1 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third inservice inspection interval (fourth inservice inspection interval for two nozzles identified in Section 1 above), currently scheduled for 2011. EGC currently proposes to perform the required volumetric examinations during the fall 2015 refueling outage concurrent with other scheduled in-vessel examinations that will occur at that time. The 2015 outage is three outages earlier than that provided to the Staff in the PWROG letter OG-06-356 (Reference 2) and is a deviation to the PWROG letter. This deviation is needed so that license renewal commitments (Reference 3) to perform inspections of the reactor vessel internals can be satisfied while minimizing the number of times that removal of the reactor vessel internals is needed.
In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current inspection interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 4).

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 3 of 11)
The methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Category B-A and B-D welds based on a negligible change in risk is contained in WCAP-16168-NP-A, Revision 2 (Reference 5). This methodology was used to develop a pilot plant analysis for Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel designs and is an extension of the work that was performed as part of the NRC's Pressurized Thermal Shock (PTS) Risk Re-Evaluation (Reference 6). The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in WCAP-16168-NP-A, Revision 2, are identified in Table 1. By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the TMI, Unit 1 reactor vessel is acceptable as shown in Table 1 below.
Table 1           Critical Parameters for Application of the Bounding Analysis Additional Plant Specific     Evaluation Parameter                        Pilot Plant Basis                   Basis         Required?
Dominant Pressurized           NRC PTS Risk Study (Reference 6)      PTS Generalization         No Thermal Shock (PTS)                                                   Study (Reference 7)
Transients in the NRC PTS Risk Study are applicable Through Wall Cracking          4.42E-07 Events per year (Reference    1.81E-12 events per       No Frequency (TWCF)               5)                                     year (calculated per Reference 5)
Frequency and Severity of      12 heatup/cooldowns per year           Bounded by 12           Yes (As Design Basis Transients        (Reference 5)                          heatup/cooldowns      required by per year             Reference 5 for B&W plants. See Appendix A.)
Cladding Layers                Single Layer (Reference 5)             Single Layer              No (Single/Multiple)

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 4 of 11)
Additional information relative to the TMI, Unit 1 reactor vessel inspections is provided in Table 2. This information confirms that satisfactory examinations have been performed on the TMI, Unit 1 reactor vessel.
Table 2          Additional Information Pertaining to the Reactor Vessel Inspections Inspection methodology:            Past inspections have been performed per the requirements of Regulatory Guide 1.150 (Reference 8) and ASME Section XI. The reactor vessel beltline region welds have also been inspected in accordance with Appendix VIII of ASME Section XI. Future inspections will be in accordance with Appendix VIII of Section XI and/or other equivalent requirements approved by the NRC.
Number of past inspections:        Two (2) inspections have been performed to date with the exception that three (3) examinations have been completed on two nozzles identified in Section 1 above.
Number of indications found:       26 recordable indications were detected in the reactor vessel beltline region during the most recent inservice inspection. All indications were acceptable per Table IWB-351 0-1 of Section XI. Only one indication is in the inner 1" of the reactor vessel thickness. This indication has a through-wall extent of 0.14". Based on the area of plate inspected in the beltline region, 42 flaws of this size would be acceptable per the proposed PTS rule (Reference 9).
Proposed inspection schedule       The next 10-year inservice inspection is currently required to be for balance of plant life:         completed no late than the 2011 refueling outage. This inspection will be performed in the 2015 refuelinQ outaQe.

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 5 of 11)
Table 3 provides additional information relative to the calculation of the TWCF for TMI, Unit 1.
Table 3            Details of TWCF Calculation for 60 EFPY of Operation Inputs Reactor Coolant System Temperature, TRcs[OF]:     I   N/A I             T wall [inches]:    I     8.565 Un-                   19 Fluence [10 Material/Flux       Cu        Ni      CF         Irradiated                 2
 #     Region/Component Description                                                                       Neutron/cm ,
Type          [wt%]    [wt%]     [OF]         RTNDT(u)
E>1 MeV]
[OF]
1      Nozzle Belt Forging ARY-59         A508 CI. 2        0.08     0.72      51                3           2.118 2       Upper Shell Plate C2789-1       SA-302 Gr BM        0.09     0.57      58                1           2.273 3       Upper Shell Plate C2789-2       SA-302 Gr BM        0.09     0.57      58                1           2.273 4       Lower Shell Plate C3307-1       SA-302 Gr BM        0.12     0.55      82                1           2.274 5       Lower Shell Plate C3251-1       SA-302 Gr BM        0.11     0.50      73                1           2.274 Nozzle Belt to Upper Shell 6                                        ASAlLinde 80       0.32     0.58    199.3           -31.1           2.118 Circ. Weld WF-70 7      Upper Shell Axial Weld WF-8       ASAlLinde 80       0.19     0.57     167            -47.6           1.513 8     Upper Shell Axial Weld WF-8        ASAlLinde 80       0.19     0.57     167            -47.6           1.513 Upper to Lower Shell Circ.
9                                       ASAlLinde 80        0.34     0.68    220.6           -74.3           2.208 WeldWF-25 Lower Shell Axial Weld SA-10                                       ASAlLinde 80        0.34     0.68   220.6            -74.3           1.358 1526 Lower Shell Axial Weld SA-11                                       ASAlLinde 80       0.34      0.68   220.6            -74.3           1.358 1526 Outputs Methodology Used to Calculate L'>.T 30:  I             Regulatory Guide 1.99, Revision 2 Controlling 19 Material                  Fluence [10 RTMAX-xx                  2    Fluence Region #                  Neutron/cm ,                       L'>.T30 [OF]  TWCF 95 -XX
[R]                        Factor (From                      E>1 MeV]
Above)
Axial Weld - AW                  10, 11       624.75           1.358         1.085             239.36       3.92E-13 Circumferential Weld - CW           6         668.56           2.118         1.204             239.97       2.90E-13 Plate - PL                          4         560.92           2.273         1.222             100.23       6.92E-14 Forging - FO                        1         524.10           2.118         1.204              61.41       1.87E-15 TWCF95-TOTAL (U.AWTWCF95-AW + u.PLTWCF95-PL + u.cw TWCF95-CW + UFoTWCF95-FO):                 1.81E-12

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR 50.55a(a)(3)(i)
(Page 6 of 11)
: 6. Duration of Proposed Alternative EGC proposes to perform the required volumetric examinations during the fall 2015 refueling outage concurrent with other scheduled in-vessel examinations that will occur at that time. This request is for the remainder of the current third interval, and subsequent 10-year lSI interval.
: 7. Precedents
: 1. "Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Evaluation of Relief Request (ISIR-29) to Extend the Third 10-Year Inservice Inspection (lSI) Interval for Reactor Vessel Weld Examination (TAC MD9934)," dated June 8, 2009 (ML091260163).
: 2. "Safety Evaluation For Relief Requests ISI-020 & 021 Reactor Vessel Weld Examination Extension - Calvert Cliffs Nuclear Power Plant, Unit No.2 (TAC Nos. MD9773 AND MD9774)," dated April 8, 2009 (ML090920077).
: 3. "Palisades Plant - Evaluation of Relief Request to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examination (TAC NO.
MD9265)," dated February 11,2009 (ML090120896).
: 4. "R.E. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC NO. MD9962)," dated July 31, 2009 (ML092080229).
: 8. References
: 1. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1995 Edition with Addenda through 1996.
: 2. OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUHP 5097-99, Task 2059," October 31, 2006.
: 3. "Safety Evaluation Report Related to the License Renewal of Three Mile Island Nuclear Station, Unit 1," Appendix A: Long Term Commitments for License Renewal of TMI-1, Commitment No. 36, dated June 2009.
: 4. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.
: 5. Pressurized Water Reactor Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated June 2008.

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR SO.SSa(a)(3)(i)
(Page 7 of 11)
: 6. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 1,2007.
: 7. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.
: 8. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," February 1983. Used for prior examinations.
: 9. SECY-07-0104, "Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," June 25, 2007 (ML070570141).

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR SO.SSa(a)(3)(i)
(Page 8 of 11)
Appendix A: Assessment of Design Basis Transients WCAP-16168-NP-A, Revision 2, states the following:
        "Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all its design basis transients and (b) identify the design basis transients that contribute to significant fatigue crack growth."
In accordance with this requirement, an evaluation was performed of the TMI, Unit 1 design basis transients. The listing of design basis transients for TMI, Unit 1 is shown in Table A-1.
Table A-1: TMI, Unit 1 Reactor Vessel Design Basis Transients for 40 Years Transient Description                                          # of Evaluated Cycles Normal Operating Transients Heatup from 70&deg;F to 8% Full Power and Cooldown from 8% Full Power                            240 Power ChanQes between 0% and 15% Power                                                      1440 Power LoadinQ and UnloadinQ between 8% and 100% Power                                      18,000 10% Step Load Increase and Decrease                                                         8,000 Feed and Bleed Operations                                                                   4,000 Test Transients - High Pressure Injection System and Core Flooding Check                     288 Valve Steam Generator FillinQ, DraininQ, FlushinQ and CleaninQ                                     540 Upset Transients Step Load Reduction                                                                          310 Reactor Trip                                                                                 400 Rapid Depressurization                                                                       40 Change of Flow                                                                               20 Rod Withdrawal Accident                                                                      40 Control Rod Drop                                                                             40 Loss of Station Power                                                                        40 Loss of Feedwater to One Steam Generator                                                     20 Loss of Feedwater Heater                                                                     40 Emergency Transients Stuck Open Turbine Bypass Valve                                                              10 Faulted Transients Steam Line Failure                                                                            1 Loss of Coolant                                                                               1 Tests Hydrotest                                                                                    20 Hot Functional TestinQ                                                                        1

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR SO.SSa(a)(3)(i)
(Page 9 of 11)
Previous flaw tolerance evaluations of the reactor vessel have shown that only the following design basis transients generate large enough changes in stress and/or occur with enough frequency to contribute to fatigue crack growth:
* Cooldown from 8% full power
* Rapid depressurization
* Power changes between 15% to 0% power
* Power loading and unloading between 8% and 100% power
* Loss of feedwater to one steam generator
* Rod withdrawal accident
* Control rod drop To determine whether the TMI, Unit 1 reactor vessel was bounded by the pilot plant analysis, the surface breaking flaw density from the B&W pilot plant "10 Year lSI Only" case (Page K-3 of WCAP-16168-NP-A, Revision 2) was compared against the surface breaking flaw density calculated based on the TMI, Unit 1 design basis transients identified above as contributors to fatigue crack growth. The following process, which is consistent with the methods described on pages 3-14 to 3-16 of WCAP-16168-NP-A, Revision 2, was used to make this comparison:
* The pressure, temperature, and film coefficient versus time histories for each transient were input into the FAVLOAD Module of the FAVOR Code along with TMI, Unit 1 specific dimensions of the reactor vessel. The FAVLOAD module generated stress intensity factors as a function of flaw depth for input to the PROBSBFD Code.
* The PROBSBFD Code was used to determine the average surface flaw density resulting from the design basis transients analyzed with the FAVLOAD module.
One run was performed for each of the design basis transients. The PROBSBFD Code was used to generate the surface breaking flaw density distribution files used in the WCAP-16168-NP-A, Revision 2. The following were used as inputs to the PROBSBFD Code:
o  The initial flaw depth was set to the TMI, Unit 1 cladding thickness rounded up to the nearest 1% of the total vessel wall thickness. To determine this depth, a cladding thickness of 0.188" was used in combination with a base metal thickness of 8.44". This combination was used because it results in the deepest initial flaw depth as a percentage of the wall thickness. This equated to 2.18%, which was rounded up to 3% and was consistent with the initial flaw depth used for the B&W pilot plant. PROBSBFD inputs for the B&W pilot plant can be found in Appendix K of WCAP-16168-NP-A, Revision 2.
o   The lSI input was chosen such that one lSI was performed at 10 years and no subsequent lSI's were performed. This corresponded to the "10 Year lSI Only" case evaluated in the WCAP pilot plant analyses.

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR SO.SSa(a)(3)(i)
(Page 10 of 11) o   The number of transients per year was chosen based on the design number per year over a 40 year plant life as identified in Table A-1.
However, consistent with the pilot plant analyses, the fatigue crack growth analysis was performed for 80 calendar years, or for twice the number of transients in Table A-1.
o   All other inputs were consistent with those used in WCAP-16168-NP-A, Revision 2.
* The average of the 1000 surface breaking flaw densities, which is output by PROBSBFD, for each design basis transient and, for their respective number of cycles per year, were added together to obtain a total average surface breaking flaw density distribution. This was done by combining the density at each specific amount of flaw growth, for each aspect ratio, for each transient.
The total average surface breaking flaw density distribution was then compared to the average distribution used in the B&W pilot plant evaluation (Page K-3 of WCAP-16168-NP-A, Revision 2) at each specific amount of flaw growth for each aspect ratio. As shown in Table A-2, the TMI, Unit 1 surface breaking flaw densities are bounded by the values in the B&W pilot plant distribution. Therefore, it can be concluded that the fatigue crack growth of 12 heat-up/cool-down transients per year used in the PWROG fatigue analysis bounds the fatigue crack growth for the TMI, Unit 1 reactor vessel design basis transients.

Relief Request RR-09-01 for the Deferral of Reactor Vessel Weld Exams in Accordance with 10 CFR SO.SSa(a)(3)(i)
(Page 11 of 11)
Table A Comparison of TMI, Unit 1 and B&W Pilot Plant Average Surface Breaking Flaw Density Distributions
("Y" Indicates TMI Flaw Density Bounded by Pilot Plant Flaw Density)
Aspect Ratio (lIa)
Flaw Growth (aft)              2               6               10               99
         +0.25%               Y               Y               Y                 Y
         +0.50%               Y               Y               Y                 Y
         +0.75%               Y               Y               Y                 Y
          +1.0%               Y               Y               Y                 Y
         +1.25%               Y               Y               Y                 Y
         +1.50%               Y               Y               Y                 Y
         +1.75%               Y               Y               Y                 Y
         +2.00%              Y                Y               Y                 Y
         +2.25%              Y                Y               Y                 Y
         +2.50%              Y                Y               Y                 Y
         +2.75%              Y                Y               Y                 Y
         +3.00%              y                y               y                 y It should be noted that the TMI, Unit 1 transients have been evaluated for license renewal and it has been determined that the actual number of transient occurrences is not expected to exceed the corresponding number in the design basis duty cycle before the end of the period of extended operation. Furthermore, TMI, Unit 1 has set reduced administrative limits on the design basis transients to meet requirements with regards to fatigue on plant components other than the reactor vessel. The TMI, Unit 1 fatigue management program will track these transient occurrences to ensure that these administrative limits are not violated. Given these administrative limits, and the fact that the TMI, Unit 1 fatigue crack growth analysis was done for 80 calendar years, the analysis discussed above is conservative.

Attachment 2 Relief Request RR-09-02

Relief Request RR-09-01 for the Deferral of Examination Category B-N-2 and B-N-3 Visual Exams in Accordance with 10 CFR SO.SSa(a)(3)(ii)
(Page 1 of 4)
: 1. ASME Code Component(s) Affected These examination categories and item numbers are from Table IWB-2500-1 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code, Section XI:
Examination Category         Item No.                    Description B-N-2                        B13.50                      Interior Attachments Within Beltline Region B-N-3                        B13.70                      Core Support Structure (Throughout this request the above examination categories are referred to as ''the sUbject examinations" and the ASME B&PV Code, Section XI is referred to as "the Code".)

===2. Applicable Code Edition and Addenda===
ASME B&PV Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1995 Edition with Addenda through 1996 (Reference 1).

===3. Applicable Code Requirement===
In accordance with IWA-2430(d)(1), each inspection interval may be reduced or extended by as much as one year. Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals. Additionally, Table IWB-2500-1, Examination Categories B-N-2 and B-N-3, Item Numbers B13.50 and B13.70 require visual examination of the accessible interior attachment welds within and beyond the beltline region and a visual examination of the accessible core support structure surfaces of the Reactor Pressure Vessel (RPV) once each 10-year interval.
The TMI, Unit 1 third 10-year Inservice Inspection (lSI) interval is scheduled to end April 19,2011; however, TMI will be applying Section XI, IWA-2430(d)(1) to extend the current interval through the fall, 2011 refueling outage.

===4. Reason for Request===
In Pressurized Water Reactor Owners Group (PWROG) Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the PWROG provided the technical and regulatory basis for decreasing the frequency of inspections by extending the ASME B&PV Code, Section XI lSI interval from the current 10 years to 20 years for Examination Categories B-A and B-D RPV welds. The U.S. Nuclear Regulatory Commission approved the topical report by letter dated May 8,2008 (Reference 3). To implement the change presented in Reference 2, Exelon Generation Company, LLC (EGC) is submitting RR-09-01, in accordance with the Safety Evaluation (Reference 3) to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. In RR-09-01, EGC identified 2015 as the year in which future inspections of the Examination Categories B-A and B-D RPV welds are scheduled to be performed. The

Relief Request RR-09-01 for the Deferral of Examination Category B-N-2 and B-N-3 Visual Exams in Accordance with 10 CFR 50.55a(a)(3)(ii)
(Page 2 of 4) intent of this relief request (RR-09-02) is to allow deferral of the subject examinations to the same time (2015 refueling outage) as the Examination Categories B-A and B-D RPV welds described in RR-09-01 .
During the 10-year lSI of the RPV shell, lower head, and nozzle welds in 2001, TMI, Unit 1 also performed visual examinations of the RPV interior attachments and the core support structure. Since the core support structure (called a core support assembly on B&W designed plants) requires removal to facilitate examination of the RPV shell, lower head, and nozzle welds, the visual examinations of Examination Categories B-N-2 and B-N-3 have historically been performed during the same outage at the end of the lSI interval.
TMI, Unit 1 has also committed to the development and implementation of a plant specific Reactor Vessel Internals (RVI) inspection program and subsequent submittal to the U.S. Nuclear Regulatory Commission two years prior to the period of extended operation (Reference 5). TMI, Unit t may elect to perform the enhanced examinations for the RVI inspection program coincident with the core barrel removal in 2015. To complete the full scope of the RVI examination it is expected to require a complete core offload and removal of all internals to facilitate implementation of the examinations.
Portions of the RVI inspection may be performed prior to this time as may be prescribed in that program.
Performing examinations related to core barrel removal during the same refueling outage will result in significant savings in dose and outage duration since the same equipment and personnel used for visual and volumetric examination of the RPV shell welds and nozzle welds from the RPV interior can be used to implement the required RVI examinations. Additionally, removing the RPV internals only once to accommodate all the examinations discussed in this relief request will result in significant savings in radiation exposure.
: 5. Proposed Alternative and Basis for Use The third ten-year lSI interval for TMI, Unit 1 began on April 20, 2001 and is scheduled to conclude on April 19, 2011. With the allowed one-year extension of IWA-2430(d)(1),
the interval may be extended to April 19, 2012.
TMI, Unit 1 proposes to perform the SUbject Examination Category B-N-2 and B-N-3 examinations for the third 10-year lSI interval prior to exiting the fall 2015 refueling outage. Completing examinations no later than the 2015 refueling outage also assures compliance with License Renewal Commitments to perform RVI examinations. MRP-227 ("Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines", Revision 0, December 2008 (Reference 4>> requires the RVI examinations to be completed within two refueling outages from the beginning of the period of extended operation (by fall 2017). The proposed alternative inspection schedule would enable the subject examinations to be performed during the 2015 refueling outage with the risk-informed extension of the RV lSI and other reqUired inspections. In accordance with 10 CFR 50.55a(a)(3)(ii), this alternative is requested on the basis that performing the examination of the RPV interior attachments and core support structure separate in time from the RPV shell, lower head, and nozzle welds

Relief Request RR-09-01 for the Deferral of Examination Category S-N-2 and S-N-3 Visual Exams in Accordance with 10 CFR 50.55a(a)(3)(ii)
(Page 3 of 4) would result in hardship or unusual difficulty without a compensating increase in quality or safety.
The examinations required by Examination Categories B-N-2 and B-N-3 require the removal of all the fuel and the core barrel from the RPV. An unnecessary risk is created by removal of the core barrel to perform a visual examination without a compensating increase in quality or safety. Further, the radiation exposure to establish the conditions for and perform the Examination Categories B-N-2 and B-N-3 examinations would essentially double if the subject examinations were performed separate in time from the RPV shell, lower head, and nozzle weld examinations.
The ASME B&PV Code, Section XI visual examinations of the RPV interior attachments and the core support assembly have been performed several times at TMI, Unit 1 with no relevant indications noted during the examinations. The examinations were last performed during the 2001 refueling outage with acceptable results.
As stated in Reference 2, "... it must be recognized that all reactor coolant pressure boundary failures occurring to date have been identified as a result of leakage, and were discovered by visual examination. The proposed RV lSI interval extension does not alter the visual examination interval. The reactor vessel would undergo, as a minimum, the Section XI Examination Category B-P pressure tests and visual examinations conducted at the end of each refueling before plant start-up, as well as leak tests with visual examinations that precede each start-up following maintenance or repair activities." The visual examinations discussed in Reference 2 are not the subject examinations (Le., B-N-2 and B-N-3) of this relief request. During the 2011 and 2013 refueling outages, TMI, Unit 1 will perform the required Examination Category B-N-1 visual examinations. These examinations will include the space that is made accessible for examination by the removal of components during normal refueling outages. These examinations are required once each period and will provide reasonable assurance of structural integrity.
As discussed further in Reference 2, defenses against human errors are preserved with the increase in inspection interval. Specifically, the increase in the inspection interval reduces the frequency for which the RV lower internals need to be removed thereby reducing the possibility for human error and damage to the core support assembly.
Therefore, in accordance with 10 CFR 50.55a(a)(3)(ii), this change in inspection timing from 2011 to 2015 for the subject examinations is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
: 6. Duration of Proposed Alternative EGC currently proposes to perform the required B-N-2 and B-N-3 examinations during the fall 2015 refueling outage concurrent with other scheduled in-vessel examinations that will occur at that time. RVI examinations in outages subsequent to the 2015 refueling outage will be in accordance with the ASME B&PV Code, Section XI, and the TMI, Unit 1 RVI inspection program.
This request is for the remainder of the current third interval, and the next 10-year inservice inspection interval.

Relief Request RR-09-01 for the Deferral of Examination Category B-N-2 and B-N-3 Visual Exams in Accordance with 10 CFR SO.SSa(a)(3)(ii)
(Page 4 of 4)
: 7. Precedents
: 1. "Safety Evaluation For Relief Requests ISI-020 & 021 Reactor Vessel Weld Examination Extension - Calvert Cliffs Nuclear Power Plant, Unit No.2 (TAC Nos. MD9773 AND MD9774)," dated April 8, 2009 (ML090920077).
: 2. "Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Evaluation of Relief Request (ISIR-30) To Extend the Third 10-Year Inservice Inspection (lSI) Interval for Visual Examination of the Reactor Pressure Vessel Interior Attachments Beyond the Beltline Region and Core Support Structure (TAC NO. ME0770)," dated June 8, 2009 (ML091320549).
: 8. References
: 1. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1995 Edition with Addenda Through 1996.
: 2. Pressurized Water Reactor Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated June 2008.
: 3. Letter from H. K. Nieh (NRC) to G. Bischoff (PWROG), dated May 8, 2008, "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG)
Topical Report (TR) WCAP-16168-NP, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (TAC No. MC9768)"," dated May 8,2008.
: 4. MRP-227 "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," Revision 0, December 2008.
: 5. "Safety Evaluation Report Related to the License Renewal of Three Mile Island Nuclear Station, Unit 1," Appendix A: Long Term Commitments for License Renewal of TMI-1, Commitment No. 36, dated June 2009.}}