DCL-11-023, Response to Summary of Telephone Conference Call Held on 02/28/2011, Between U.S. Nuclear Regulatory Commission & PG&E Co. Concerning Responses to Requests for Additional Info. for Renewal Application
| ML110940188 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  (DPR-080, DPR-082) |
| Issue date: | 03/25/2011 |
| From: | Becker J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-11-023 | |
| Download: ML110940188 (8) | |
Text
- Pacific Gas and Electric Company-James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P 0. Box 56 Avila Beach, CA 93424 805.545.3462 March 25, 2011 Internal: 691.3462 Fax: 805.545.6445 PG&E Letter DCL-1 1-023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to Summary of Telephone Conference Call Held on February 28, 2011, Between the U.S. Nuclear Regulatory Commission and Pacific Gas and Electric Company Concerning Responses to Requests for Additional Information for the Diablo Canyon License Renewal Application
Dear Commissioners and Staff:
By Pacific Gas and Electric Company (PG&E) Letter DCL-09-079, "License Renewal Application," dated November 23, 2009, PG&E submitted an application to the U.S.
Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the Applicant's Environmental Report -
Operating License Renewal Stage.
A telephone conference between the NRC and representatives of PG&E was held on February 28, 2011, to obtain clarification on the applicant's response to request for additional information (RAI) submitted to the NRC in PG&E Letters DCL-1 0-168, "Response to NRC Letter dated December 20, 2010, Request for Additional Information (Set 37) for the Diablo Canyon License Renewal Application," dated January 7, 2011 and DCL-10-167, "Response to NRC-Letter dated. December 20, 2010, Request for Additional Information (Set 36) for the Diablo Canyon License Renewal Application," dated January 12, 2011, regarding Time Limited Aging Analysis (TLAA) and aging management programs. The telephone conference call was useful in clarifying the intent of PG&E's responses. PG&E agreed to supplement the previous responses.
In addition, clarifying information was obtained regarding applicable transients for DCPP's Model 93A Reactor Coolant Pump (RCP) casings. These changes are shown in Enclosure 3.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Caltlaway e Comanche Peak
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Document Control Desk March 25, 2011 Page 2 PG&E Letter DCL-1 1-023 PG&E's supplemental information to the RAI responses for which the staff requested clarification is provided in Enclosure 1. LRA Amendment 42 resulting from the responses is included in Enclosure 2 showing the changed pages with line-in/line-out annotations. PG&E amends commitments in revised LRA Table A4-1, License Renewal Commitments, shown in Enclosure 2.
If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4160.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 25, 2011.
James R. Bfcke Site Vice President tlg/50378804 Enclosures cc:
Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial B. Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Fred Lyon, NRC Project Manager, Office of Nuclear Reactor Regulation Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, License Renewal A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway e Comanche Peak e Diablo Canyon
- Palo Verde e San Onofre
- South Texas Project
- Wolf Creek PG&E Letter DCL-1 1-023 Page 1 of 2 Response to Telephone Conference Call Held on February 28, 2011, Between the U.S..Nuclear Regulatory Commission and Pacific Gas and Electric Company Concerning Responses to Requests for Additional Information for the Diablo Canyon License Renewal Application RAI 4.1-7 Follow-up The Staff requested clarification in Diablo Canyon Power Plant (DCPP) Commitment 22 regarding validation of the baffle and former bolt inspection interval under the Reactor Vessel Internals Aging Management Program.
PG&E Response to RAI 4.1-7 Follow-up As stated in PG&E's response to RAI 4.1-7 in PG&E Letter DCL-10-167, dated January 12, 2011, fatigue of the baffle and former bolts will be managed in accordance with the Reactor Vessel Internals Aging Management Program. The schedule for inspection of the baffle and former bolts will be validated on a plant-specific basis to ensure that it will appropriately manage the design fatigue analysis. See amended license renewal application (LRA) Commitment 22 in Enclosure 2.
RAI 4.3-15 Follow-up Request 1:
The Staff requested clarification in' DCPP Commitment 58 on which guidance will be used to evaluate more limiting components for the effects of the reactor coolant environment.
Request 2:
The staff requested confirmation that DCPP maintains dissolved oxygen (DO) level below 0.05 ppm and has never had a prolonged period of DO greater than 0.05 ppm.
PG&E Response to RAI 4.3-15 Follow-up Request 1:
In accordance with Commitment 58, if more limiting components are identified during the review of design basis ASME Class 1 component fatigue evaluations, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage in using the Metal Fatigue of Reactor Coolant Pressure Boundary program. The effect of the reactor coolant environment on DCPP fatigue usage will be evaluated using material-specific guidance presented in NUREG/CR-6583 for carbon PG&E Letter DCL-1 1-023 Page 2 of 2 and low alloy steels, NUREG/CR-5704 for stainless steels, and NUREG/CR-6909 for nickel alloys. See amended LRA Commitment 58 in Enclosure 2.
Request 2:
DCPP has never experienced a dissolved oxygen (DO) spike exceeding 0.05 parts per million (ppm) in the reactor coolant system (RCS) during operation. During operation, Unit 1 and Unit 2 remain less than 0.002 ppm DO. Elevated hydrogen levels prevent DO from exceeding 0.002 ppm. The RCS water is sampled three times per week for hydrogen and four times per week for DO during operation.
PG&E Letter DCL-11-023 Page 1 of 2 LRA Amendment 42 LRA Section RAI Table A4.1 4.1-7 Table A4.1 4.3-15 PG&E Letter DCL-11-023 Page 2 of 2 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1 License Renewal Commitments Item #
Commitment LRA Implementation Section Schedule 22 B. For Reactor Vessel Internals:
4.3.3 Concurrent with industry (1) Participate in the industry programs for investigating and managing aging effects on initiatives and upon reactor internals; (2) evaluate and implement the results of the industry programs as completion submit an applicable to the reactor internals; and (3) upon completion of these programs, but not inspection plan and not less less than 24 months before entering the period of extended operation, PG&E will than 24 months before submit an inspection plan for reactor internals to the NRC for review and approval, entering the period of PG&E will validate the schedule for inspection of the baffle and former bolts on a plant-extended operation.
specific basis to ensure that it will appropriately manage the design fatigue analysis.
58 PG&E will perform a review of design basis ASME Class 1 component fatigue 4.3.4 Prior to the period of evaluations to determine whether the NUREG/CR-6260-based components that have extended operation been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the DCPP plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. The effect of the reactor coolant environment on DCPP fatigue usage will be evaluated using material-specific guidance presented in NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-5704 for stainless steels, and NUREG/CR-6909 for nickel alloys. This additional evaluation will be performed through the Metal Fatigue of Reactor Coolant Pressure Boundary Program in accordance with 10 CFR 54.21 (c)(1)(iii).
I. _____________________________________________________________________________________
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PG&E Letter DCL-1 1-023 Page 1 of 2 Summary of Table Revisions PG&E Letter DCL-10-168, dated January 7, 2011, provided Request for Additional Information (RAI) responses pertaining to the DCPP License Renewal Application.
As part of the response to RAI 4.3-1 (follow-up), PG&E provided a table of transients.
One of the transient sources was WCAP-1 3045 Flaw Growth Analysis.
Since the submittal of PG&E letter DCL-10-168, Westinghouse has provided a supplemental letter containing clarification between the transients in relation to generic Reactor Coolant Pump Casing (RCP) Model 93 used in WCAP-1 3405 and DCPP's Model 93A. The revised Table 1 shows these changed transients with line-in/line-out annotations.
PG&E Letter DCL-11-023 Page 2 of 2 Table 1 Auxiliary WCAP-13045 60-Year Transient LBB Feedwater Flaw Growth Projections Analysis n
Analysis (Unit LUnit 2)
Analysis A ayi U i /nt2 Normal Conditions RCS heatup and cooldown at 200 250 200 85/65
<100°F/hr Unit loading and unloading at 5%
18,300 Not Included Not Included Not Projected of full power/min Step increase and decrease of 2,000 Not Included 2TOONot 56 / 61 Included 10% of full power 200 Not Included 20ONot 11/9 Large step load decrease Included Steady state fluctuations 106 Not Included 3,!50,00Not Not Projected Included Upset Conditions Loss of load (above 15% full 80 NotIncluded 80 18/10 power), without immediate turbine or reactor trip 40 Not Included 40 2/ 3 Loss of all offsite power Partial loss of flow 80 Not Included Net-3 / 8 J-Rok~ied8O Reactor trip from full power 400 Not Included 3,0Not 100 / 83 Included Inadvertent RCS depressurization Not Included Not Included 2-ONot 3 / 3 Included Control rod drop Not Included Not Included 8ONot 5 / 2 Included Test Conditions Turbine roll test 10 Not Included Net-8/ 9 Pfaided2O Primary side hydrostatic test 5
Not Included Not Included 2 / 2 50 Not Included Net-5/5 Primary side leak test kiu44ded200 Cold hydrostatic test 10 Not Included 10 Not Projected Faulted Conditions 7.5M Hosgri earthquake Not Included 5
Not Included 1 / 1 Emergency Conditions Complete loss of flow Not Included Not Included 5Not Included 1 I 1