RS-14-088, Relief Request 13R-23 to Extend the Reactor Vessel Inservice Inspection Interval

From kanterella
Jump to navigation Jump to search

Relief Request 13R-23 to Extend the Reactor Vessel Inservice Inspection Interval
ML14069A559
Person / Time
Site: Byron Constellation icon.png
Issue date: 03/10/2014
From: Gullott D
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-088
Download: ML14069A559 (9)
v • d • e


Text

RS-14-088 10 CFR 50.55a March 10, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Byron Station, Unit 1 Facility Operating License No. NPF-37 NRC Docket No. STN-50-454

Subject:

Relief Request 13R-23 to Extend the Reactor Vessel lnservice Inspection Interval

References:

1) WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011
2) Letter from R. A. Nelson (NRC) toW. A. Nowinowski (PWROG), "Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding 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 July 26, 2011 In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), Exelon Generation Company, LLC, (EGC) hereby requests NRC approval of the attached relief request associated with the Third 10-year lnservice Inspection (lSI) Program Interval for Byron Station, Unit 1. The third interval of the Byron Station lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda.

NRC approval of the attached relief is requested on the basis provided in topical report WCAP-16168-NP-A, Revision 3 (Reference 1), for extending the ASME B&PV Code,Section XI lSI interval from the current 10 years to 20 years as the alternative inspection interval that provides an acceptable level of quality and safety for Examination Categories B-A and B-D reactor pressure vessel (RPV) welds for Byron Station, Unit 1. The NRC approved the topical report by letter dated July 26, 2011 (Reference 2).

The details of the 10 CFR 50.55a relief request are enclosed. EGC requests approval of this relief request by September 4, 2015 to support Byron Station, Unit 1 refueling outage (B1 R20) scheduled to commence on September 8, 2015.

March 10, 2014 U.S. Nuclear Regulatory Commission Page2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Ms. Dwi Murray at (630) 657-3695.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC

Attachment:

10 CFR 50.55a Relief Request 13R-23 cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Byron Station NRR Project Manager, Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(a)(3)(i)

Revision 0 Page 1 of 7 1.0 ASME CODE COMPONENTS AFFECTED:

The affected component is the Byron Station, Unit 1 reactor vessel (RV), specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the reactor vessel.

These affected Class 1 examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME B&PV Code,Section XI.

Examination Item Category Number Description B-A B1.11 Circumferential Shell Welds B*A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds (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.")

See Table 1 for a listing of the applicable examination areas of the Unit 1 reactor vessel.

2.0 APPLICABLE CODE EDITION AND ADDENDA:

The Third lntervallnservice Inspection (lSI) program is based on the ASME B&PV Code,Section XI, 2001 Edition through the 2003 Addenda (Reference 1).

The applicable Code for subsequent lSI intervals will be implemented in accordance with the requirements of 10 CFR 50.55a.

3.0 APPLICABLE CODE REQUIREMENT:

ASME Section XI, Paragraph IWB-2412, "Inspection Program B," requires volumetric examination of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each 10-year inspection interval. The Byron Station, Unit 1 Third 10-year lSI Interval ends on July 15, 2016.

4.0 REASON FOR REQUEST:

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety. 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 10-year interval.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R~23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(a}(3}(i)

Revision 0 Page 2 of7 Extension of the interval between examinations of specific Category B-A and B-D welds from 10 years to a maximum of 20 years will result in a reduction in person-rem exposure and examination costs.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

Exelon Generation Company, LLC (EGC) proposes to defer the ASME B&PV Code required volumetric examination of specific Byron Unit 1 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third lSI interval, currently scheduled to be performed in 2015. EGC proposes to perform these ASME B&PV Code required volumetric examinations during the fourth lSI interval for Byron Unit 1 in 2025. This date is consistent with those provided in Pressurized Water Reactor Owners Group (PWROG) letter OG-06-356 (Reference 2) and the latest revised implementation plan provided in PWROG letter OG-1 0-238 (Reference 3).

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current examination interval can be extended based on a negligible change in risk when compared to the risk criteria specified in Regulatory Guide 1.174 (Reference 4). The methodology used to conduct this analysis is based on that defined in WCAP-16168-NP-A, Revision 3 (Reference 5). This methodology focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Byron Station, Unit 1 were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of WCAP-16168-NP-A, Revision 3 identifies the parameters to be compared. By demonstrating that the parameters for Byron Station, Unit 1 are bounded by the results of the Westinghouse pilot plant, the methodology qualifies Byron Station, Unit 1 for an lSI interval extension.

The following Table 1 lists the applicable examination areas addressed under this relief request of the Byron Station, Unit 1 reactor vessel.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(a)(3)(i)

Revision 0 Page 3 of7 Table 1: Examination Category, Item Number, Component Description, and Component Identification of Applicable Examinations for Byron Unit 1 Component Identification (Exam Area)

Exam Item Component Description Unit 1 Vessel:

Category #

1RC*01*R B-A B1.30 Flange to Nozzle Belt (Upper Shell) WR-7 B-A B1.11 Nozzle Belt (Upper Shell) to Mid Shell WR-34 B-A B1.11 Mid Shell to Lower Shell WR-18 B-A B1.11 Lower Shell to Bottom Head Torus WR-29 B-A B1.21 Bottom Head Torus to Head Dome WR-16 Nozzle Components B-D B3.90 Outlet Nozzle @ 22° RPVN-A B-D B3.90 Inlet Nozzle @ 6r RPVN-B B-D B3.90 Inlet Nozzle @ 113° RPVN-C B-D B3.90 Outlet Nozzle @ 158° RPVN-D B-D B3.90 Outlet Nozzle @ 202° RPVN-E B-D B3.90 Inlet Nozzle @ 24r RPVN-F B-D B3.90 Inlet Nozzle @ 293° RPVN-G B-D B3.90 Outlet Nozzle @ 338° RPVN-H Table 2 below lists the critical parameters investigated in WCAP-16168-NP-A, Revision 3 and compares the results of the Westinghouse pilot plant to those of Byron Station, Unit 1.

Table 2: Critical Parameters for the Application of Bounding Analysis for Byron Unit 1 Additional Parameter Pilot Plant Basis Plant-Specific Basis Evaluation Required?

Dominant Pressurized Thermal Shock (PTS) Transients in the NRC PTS Risk PTS Generalization Study No NRC PTS Risk Study are Study (Reference 6) (Reference 7)

Applicable 1.76E-08 Events Through-Wall Cracking Frequency 2.30E-14 Events per year per year No (TWCF) (Calculated per Reference 5)

(Reference 5) 7 heatup/cooldown Bounded by 7 Frequency and Severity of Design cycles per year heatup/cooldown No Basis Transients (Reference 5) cycles per year Single Layer Cladding Layers (Single/Multiple) Single Layer No (Reference 5)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(a)(3)(i)

Revision 0 Page4of7 Table 3 below provides a summary of the latest RV inspection results for Byron Station, Unit 1 and evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on Byron Station, Unit 1 RV.

Table 3: Additional Information Pertaining to Reactor Vessel Inspection for Byron Unit 1 Inspection The latest Unit 1 lSI was conducted in accordance with the ASME Code,Section XI methodology: and Section V, 1989 Edition, no Addenda. Examinations of Category B-A and B-D welds were performed to ASME Section XI Appendix VIII, 1995 Edition with the 1996 Addenda, as modified by 10 CFR 50.55a(b}(2)(xiv, xv and xvi).

Future inservice inspections will be performed to the applicable ASME Section XI Appendix VIII and 10 CFR 50.55a requirements.

Number of past Two 10-Year inservice inspections have been performed.

inspections:

Number of There were two indications identified in the beltline region circumferential weld during indications the most recent inservice inspection. These indications are acceptable per Table found: IWB-351 0-1 of Section XI of the ASME Code. Only one indication, contained in the 1

nozzle shell forging material, is within the inner 1/10 h or 1" of the reactor vessel thickness. This indication is acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 8), since the flaw is less than the allowable number of flaws for each flaw size increment. A disposition of this flaw against the limits of the Alternate PTS Rule is shown in the table below.

Through-Wall Extent, TWE (in.) Scaled Maximum Number of forging number of forging Flaws TWEMIN TWEMAX flaws (Axiai/Circ.)

0.075 0.375 76 1 (1 /0) 0.125 0.375 30 1 {1/0) 0.175 0.375 8 1 {1/0)

Proposed The third inservice inspection for Unit 1 is currently scheduled for 2015. This inspection inspection will be performed in 2025.

schedule for This date is consistent with those provided in PWROG letter OG-06-356 (Reference 2) balance of and the latest revised implementation plan in PWROG letter OG-10-238 (Reference plant life:

3).

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a{a)(3)(i)

Revision 0 Page 5 of7 Table 4 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF) for Byron Station, Unit 1.

Table 4: Details of TWCF Calculation for Unit 1 at 57 Effective Full Power Years {EFPY)

Inputs Reactor Coolant System Temperature, T c [°F]: N/A T wall [inches]: 8.625 19 cu<a> Ni(a) R.G. CF<a> Fluence [1 0 Region and Component Material RTNDT 2 No. 1.99 [OF] (a)[oF] Neutron/cm ,

Description Heat No. [wt%] [wt%] (u)

Pas. E > 1.0 MeV]

1 Nozzle Shell Forging 123J218 0.05 0.72 1.1 31.0 30 1.15 NS Forging to IS Forging Circ.

2 442011 0.03 0.67 2.1 26.1 10 1.15 Weld WF-501 3 Intermediate Shell Forging 5P-5933 0.04 0.74 2.1 30.6 40 3.21 IS Forging to LS Forging Circ.

4 442002 0.04 0.63 2.1 66.5 -30 3.08 WeldWF-336 5 Lower Shell Forging 5P-5951 0.04 0.64 1.1 26.0 10 3.21 Outputs Methodology Used to Calculate f1T 30 : Regulatory Guide 1.99, Revision 2<bl Controlling 19 Fluence [10 FF Material Region RT MAX-XX 2 f1Tso

[OF] Neutron/cm , (Fiuence [oF] TWCF9s-xx No.

E > 1.0 MeV] Factor)

(From Above)

Limiting Forging - FO 3 79.98 3.21 1.307 39.98 9.19E-15 Limiting Circumferential 3 79.69 3.08 1.297 39.69 O.OOE+OO Weld-CW TWCF9s-TOTAL(0FoTWCF9s-Fo + OcwTWCF9s-cw): 2.30-14 Notes:

(a) Data obtained from Reference 9.

(b) Reference 10.

6.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Inspection Interval for Byron Station, Unit 1.

This request is applicable to the Byron Station, Unit 1 lSI program for the third 10-year intervals.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-23 Request for Relief for Alternative Requirements for Reactor Vessellnservice Inspection Interval In Accordance with 10 CFR 50.55a(a)(3)(i)

Revision 0 Page 6 of7

7.0 PRECEDENTS

The NRC has previously authorized similar relief requests for the extended inspection interval for reactor vessel examinations using WCAP-16168-NP-A. Authorization has been recently granted for relief requests from the following stations.

1. RR-111-07 for Virgil C. Summer Nuclear Station Unit 1 was authorized by NRC Safety Evaluation dated July 19,2012, ADAMS Accession No. ML12191A163.
2. RR-23 for H. B. Robinson Steam Electric Plant Unit 2 was authorized by NRC Safety Evaluation dated July 22, 2011, ADAMS Accession No. ML11186A011.
3. AN02-ISI-004 for Arkansas Nuclear One, Unit 2 was authorized by NRC Safety Evaluation dated September 21, 201 0, ADAMS Accession No. ML102450654.
4. RR-09-01 and RR-09-02 for Three Mile Island Nuclear Station Unit 1 were authorized by NRC Safety Evaluation dated September 21 , 201 0, ADAMS Accession No. ML102390018.
5. FNP-ISI-ALT-09 for Joseph M. Farley Nuclear Plant, Unit 2 was authorized by NRC Safety Evaluation dated July 12,2010, ADAMS Accession No. ML101750402.
6. RR-40 for Palo Verde Nuclear Generating Station Units 1, 2, and 3 was authorized by NRC Safety Evaluation dated February 22,2010, ADAMS Accession No. ML100290415.
7. S1-13R-93, S2-13R-94, and SC-13R-95 for Salem Nuclear Generating Station, Units 1 and 2 were authorized by NRC Safety Evaluation dated February 22, 201 0, ADAMS Accession No. ML100491550.
8. ISI-020 and ISI-021 for Calvert Cliffs Nuclear Power Plant Unit 2 were authorized by NRC Safety Evaluation dated April 8, 2009, ADAMS Accession No. ML090920077.

8.0 REFERENCES

1. ASME Boiler and Pressure Vessel Code,Section XI and Section 2001 Edition, with 2003 Addenda, American Society of Mechanical Engineers, New York.
2. OG-06-356, "Plan for Plant-Specific Implementation of Extended lnservice 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. OG-1 0-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended In service Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R*23 Request for Relief for Alternative Requirements for Reactor Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(a)(3)(i)

Revision 0 Page 7 of7 Informed Extension of the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," July 12, 2010.

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. WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," October 2011.
6. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010 (ADAMS Accession No. ML070860156).
7. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession No. ML042880482).
8. Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," U.S.

Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. 1, dated January 4, 201 0 and No. 22 with corrections to part (g) dated February 3, 201 0, March 8, 201 0, and November 26, 201 0.

9. WCAP-17606-NP, Revision 0, "Byron Station Units 1 and 2 Reactor Vessel Integrity Evaluation to Support License Renewal Time-Limited Aging Analysis," December 2012.
10. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
Retrieved from "https://kanterella.com/w/index.php?title=RS-14-088,_Relief_Request_13R-23_to_Extend_the_Reactor_Vessel_Inservice_Inspection_Interval&oldid=2453561"