ML110880113
| ML110880113 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/25/2011 |
| From: | Becker J R Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-11-023, OL-DPR-80, OL-DPR-82 | |
| Download: ML110880113 (8) | |
Text
Pacific Gas and Electric Company March 25, 2011 PG&E Letter DCL-11-023 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 James R. Becker Site Vice President Diablo Canyon Power Plant Mail Code 104/5/601
- p. O. Box 56 Avila Beach, CA 93424 805.545.3462 Internal:
691.3462 Fax: 805.545.6445 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-10-168, "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-1 0-167, "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 Callaway 0 Comanche Peak I) Diablo Canyon. Palo Verde f) San Onofre. South Texas Project. Wolf Creek Document Control Desk March 25, 2011 Page 2 PG&E Letter DCL-11-023 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 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. 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 0 Comanche Peak 0 Diablo Canyon. Palo Verde. San Onofre. South Texas Project
- Wolf Creek Enclosure 1 PG&E Letter DCL-11-023 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 RAJ 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-1 0-167, 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. RAJ 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 Enclosure 1 PG&E Letter DCL-11-023 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.
Enclosure 2 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 I Enclosure 2 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 NUREGICR-6583 for carbon and low alloy steels , NUREGICR-5704 for stainless steels , and NUREGICR-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).
Enclosure 3 PG&E Letter DCL-11-023 Page 1 of 2 Summary of Table Revisions PG&E Letter DCL-1 0-168, 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-13045 Flaw Growth Analysis.
Since the submittal of PG&E letter DCL-1 0-168, 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-13405 and DCPP's Model 93A. The revised Table 1 shows these changed transients with in/line-out annotations.
I I I I Enclosure 3 PG&E Letter DCL-11-023 Page 2 of 2 Transient RCS heatup and cooldown at Unit loading and unloading at 5% of full power/min Step increase and decrease of 10% of full power Large step load decrease Steady state fluctuations Loss of load (above 15% full power), without immediate turbine or reactor trip Loss of all offsite power Partial loss of flow Reactor trip from full power Inadvertent RCS depressurization Control rod drop Turbine roll test Primary side hydrostatic test Primary side leak test Cold hydrostatic test 7.5M Hosgri earthquake I Complete loss of flow Table 1 Auxiliary LBB Feedwater Analysis Line 567 Analysis Normal Conditions 200 250 18,300 Not Included 2,000 Not Included 200 Not Included 10° Not Included Upset Conditions 80 Not Included 40 Not Included 80 Not Included 400 Not Included Not Included Not Included Not Included Not Included Test Conditions 10 Not Included 5 Not Included 50 Not Included 10 Not Included Faulted Conditions Not Included 5 Emergency Conditions Not Included Not Included WCAP-13045 60-Year Flaw Growth Projections Analysis (Unit 1/Unit 2) 200 85/65 Not Included Not Projected
&,OOG N ot 56/61 I nc l uded 11 /9 I nc l uded d, §G,GGG Not Not Projected I ncluded 80 18/10 40 2/3 Net-3/8 IAGll:Jeee BO 100/83 I nc l uded 2:Q N ot 3/3 I ncluded W Not 5/2 Inc l uded Net-8/9 IAGII:Jeee 20 Not Included 2/2 Net-5/5 IAGII:Jeee 200 10 Not Projected Not Included 1 / 1 a N ot I nc lud ed 1 / 1