RNP-RA/10-0108, H. B. Robinson,Response to Request for Additional Information Regarding Changes to Technical Specification Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation, and Section 3.3.6, Containment Ventilation Isolation

H. B. Robinson,Response to Request for Additional Information Regarding Changes to Technical Specification Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation, and Section 3.3.6, Containment Ventilation Isolation Instr

ML103090167
Person / Time
Site:	Robinson
Issue date:	11/01/2010
From:	Bernie White Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/10-0108
Download: ML103090167 (6)

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2 Progress Energy Serial: RNP-RA/10-0108 NOV 0 12010 United States Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, Maryland 20852 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING CHANGES TO TECHNICAL SPECIFICATION SECTION 3.3.2, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION, AND SECTION 3.3.6, CONTAINMENT VENTILATION ISOLATION INSTRUMENTATION (TAC NO. ME3507)

Ladies and Gentlemen:

By letter dated July 20, 2010, the NRC requested that Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), respond to a request for additional information (RAI) regarding the proposed license amendment request to Technical Specifications Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation, and Section 3.3.6, Containment Ventilation Isolation Instrumentation that was submitted on March 5, 2010.

Attachment I provides an Affirmation in accordance with the provisions of 10 CFR 50.30(b).

Attachment II provides the response to the NRC RAI for this license amendment request.

In accordance with 10 CFR 50.91, a copy of this application is being provided to the State of South Carolina.

If you have any questions concerning this matter, please contact Curt Castell at (843) 857-1626.

Sincerely, Beja in C. White Manager - Support Services - Nuclear RAC/rac Progress Nuclear Robinson Energy Plant Carolinas, Inc. A co l 3581 West Entrance Road Hartsville, SC 29550

United States Nuclear Regulatory Commission Serial: RNP-RA/10-0108 Page 2 of 2 Attachments: I. Affirmation II. Response To NRC Request For Additional Information Regarding Changes To Technical Specification Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation, and Section 3.3.6, Containment Ventilation Isolation Instrumentation c: Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

Mr. A. Gantt, Chief, Bureau of Radiological Health (SC)

Mr. L. A. Reyes, NRC, Region II Mr. T. Orf, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP Attorney General (SC)

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/10-0108 Page 1 of 1 AFFIRMATION The information contained in letter RNP-RA/10-0 108 is true and correct to the best of my information, knowledge, and belief; and the sources of my information are officers, employees, contractors, and agents of Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. I declare under penalty of perjury that the o egoing is true and correct.

Executed On: NOV 0 1 2010 Vice dent, J. D uncan, P II Vice Prsdent, HBRSEP, Unit No. 2

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0108 Page 1 of 3 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING CHANGES TO TECHNICAL SPECIFICATION SECTION 3.3.2, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION, AND SECTION 3.3.6, CONTAINMENT VENTILATION ISOLATION INSTRUMENTATION NRC Question 1:

This technical specifications change will effectively increase the allowable channel bypassed time from 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which increases the probability of a safety function failure. Has this change been factored into the plant's probable risk assessment? If so, then what was the overall impact?

Response

The Technical Specifications (TS) change is not modeled in the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, plant specific probabilistic safety assessment (PSA).

Unavailability in the PSA is based on plant specific data for recurring risk significant maintenance events. The PSA has not modeled the allowed outage time as unavailability.

Therefore, the proposed TS revision to allow additional channel bypass time is not evaluated in the PSA.

The Containment Pressure High High signal containment spray actuation function is modeled in the PSA Large Early Release Frequency (LERF) model. The other two safety functions (MSIV closure and Containment Phase B isolation) are not modeled.

A bounding analysis is performed by setting one containment pressure channel as failed in the PSA model (Calculation RNP-F/PSA-0077, Rev. 0) and quantifying the LERF results.

The base LERF value is 3.89E-7/yr and the failed channel LERF value is 3.92E-7/yr. For a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AOT extension this results in the following change in Large Early Release Probability (LERP), which as shown below is a very small risk increase.

Delta LERP = (3.92E-7-3.89E-7)*6hrs/8760hrs = 2E-12 per use of the allowed bypass time extension.

NRC Question 2:

In the justification for the proposed change provided, the licensee stated that the added 6-hour allowance is acceptable based on the low probability of an accident or transient occurring during that time period. It also states that there is a high probability that the channels will still perform their actuation function. The staff requests that the licensee provide quantifiable bases for these statements.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0108 Page 2 of 3

Response

The reliability of the unaffected channels is considered high based upon industry data such as NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at US Commercial Nuclear Power Plants," February 2007, which provides recent industry failure rates for these types of components. The HBRSEP, Unit No. 2, plant specific PSA, using industry data, has reliability of better than 99% per channel. The results of an evaluation of the probability of an applicable event during the six hour extended time determined the increase in risk is negligible at approximately 5E-09, as shown below.

The low probability of an event during the requested six hour extended time is determined as follows using a probabilistic approach. The event can be quantified as follows:

APf(Event) = P(IE) x APf(Signal)

The variables are described below:

P(IE) - Probability of initiating event:

The initiating events that will actuate the Containment Pressure High High signal are Loss of Coolant Accidents and secondary line breaks inside containment. The respective frequencies of these events are 5E-4 and 1E-3 per year, as taken from NUREG/CR-5750, "Rates of Initiating Events at U.S. Nuclear Power Plants: 1987- 1995."

The probability of these initiating events occurring during the requested extension time is as follows:

P(IE) -- (6 hrs/8760 hrs/yr) x (1.5E-3/yr) = 1E-6 APf(Signal) - Increase in probability of signal failure The Containment Pressure High High signal is based on a 2-of-3 taken twice logic. Bypass of a single channel reduces the logic to 2-of-2 for the affected train. The failure probability used in the PSA of each channel is approximately 2.5E-3 (inclusive of the channel components and power supplies). The resultant increase in failure probability of bypassing a channel for the requested extension time is the difference between the bypass condition failure probability (any one of two channels failing) and the base condition failure probability (three possible combinations of any two channels failing) as follows:

APf(signal) = Pf(LCO) - Pf(base) = (2 x 2.5E-3) - (3 x (2.5E-3 x 2.5E-3))

APf(signal) = 5E-3 Common cause that only applies to the base condition is conservatively omitted, as it would only diminish the overall increase in calculated risk.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0108 Page 3 of 3 APf(Event) - Increase Probability of signal failure when needed Combining these two probabilities gives the increase in likelihood of having an initiating event with failure of the Containment Pressure High High actuation signal:

APf(Event) = 1E-6 x 5E-3 = 5E-9 NRC Question 3:

The licensee stated that in most cases, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> should provide sufficient time to perform maintenance activities required to return a channel to service. If this is the case, then why is the current 6-hour completion time not sufficient to support the necessary maintenance activities?

The affected channel would not be required to be tripped during this initial time period and module replacements could be performed without necessitating the use of the extraordinary measures described in the application.

Response

It is expected that in the majority of cases, such as pre-planned maintenance, activities can be completed in the current six hour completion time as currently allowed. However, as experienced in the June 29, 2009, event that prompted the need for this license amendment request, emergent conditions can pose a challenge to meeting the six hour criterion. On that date, the time it took for planning, troubleshooting, and obtaining replacement parts took greater than six hours. For other safeguard channels this would not be a problem, as the channel can remain de-energized and in the trip condition while the channel is repaired. However, because of the unique design of the Containment Pressure High High signal (must be energized to trip), it is desired that the channel be de-energized and removed from the trip condition during repair. Six additional hours is considered sufficient time for these repair activities.