ML17262A089

Request for Additional Information Concerning Permanent Extension of Type a and Type C Leak Rate Test Frequencies (RS-17-051)

ML17262A089
Person / Time
Site:	Quad Cities
Issue date:	09/19/2017
From:	Kimberly Green Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Green K, NRR/DORL/LPLIII, 415-1627
References
CAC MF9675, CAC MF9676
Download: ML17262A089 (5)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 19, 2017 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1AND2- REQUEST FOR ADDITIONAL INFORMATION CONCERNING PERMANENT EXTENSION OF TYPE A AND TYPE C LEAK RATE TEST FREQUENCIES (RS-17-051)

(CAC NOS. MF9675 AND MF9676)

Dear Mr. Hanson:

By letter dated April 27, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17121A449), Exelon Generation Company, LLC (EGC) submitted a license amendment request for Quad Cities Nuclear Power Station, Units 1 and 2. The proposed amendment would modify Technical Specification 5.5.12, "Primary Containment Leakage Rate Testing Program," to allow for the permanent extension of the Type A Integrated Leak Rate Testing and Type C Leak Rate Testing frequencies.

The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and response to the NRC staff's request for additional information (RAI) and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. A draft RAI was transmitted by email to Mr. Ken Nicely on September 14, 2017. A clarification call was held on September 18, 2017. No changes to the draft RAI were needed as a result of the clarification call. Based on a discussion with Mr. Ken Nicely, it was agreed that EGC will provide a response to the RAI within 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.

B. Hanson If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1627.

Sincerely, K\k~

Kimberly J. Green, enior Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION EXELON GENERATION COMPANY. LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1AND2 DOCKET NOS. 50-254 and 50-265 By letter dated April 27, 2017, Exelon Generation Company, LLC (EGC), submitted a license amendment request (LAR) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17121A449). The proposed amendment would modify Technical Specification 5.5.12, "Primary Containment Leakage Rate Testing Program," to allow for the permanent extension of the Type A Integrated Leak Rate Testing and Type C Leak Rate Testing frequencies. By letter dated July 17, 2017 (ADAMS Accession No. ML17198A229),

EGC supplemented the LAR with additional information in response to the U.S. Nuclear Regulatory Commission (NRC) staff's requests (ADAMS Accession Nos. ML17180A153 and ML17198A229). The NRC staff has reviewed the additional information and has determined that further information below is needed to support the NRC staff's continued technical review of the LAR.

RAI 5-A Section 4.2.6 of EPRI [Electric Power Research Institute] TR [Topical Report)-1009325, Revision 2-A, states that "[p]lants that rely on containment overpressure for net positive suction head (NPSH) for emergency core cooling system (ECCS) injection for certain accident sequences may experience an increase in CDF [core damage frequency)," therefore requiring a risk assessment. In response to request for additional information (RAI) 5, EGC described the probabilistic risk assessment (PRA) modeling to estimate the change in risk from loss of NPSH to the ECCS pumps and provided a LiCDF estimate of 2.4E-8/year.

a. As described in the RAI response, the risk analysis appears to assume a total loss of containment heat removal for all accident scenarios that were considered, such as transients, loss-of-coolant accident, anticipated transient without scram, special initiators, etc., which are listed in Table 5-1 of the RAI response. However no justification was provided for scenarios with containment decay heat removal available.

Explain and justify why the loss of containment overpressure impacting NPSH for the ECCS injection is not a concern in any accident scenario with containment decay heat removal available. Alternatively, if any additional accident scenarios are identified to contribute to the risk increase, provide an updated estimate of LiCDF.

b. The RAI response attempts to explain the PRA modeling for scenarios "post containment failure" and for scenarios with "successful pool venting" in Figures 5-4 and 5-5. Since the containment would already be failed due to the postulated pre-existing containment leak, further PRA modeling appears unnecessary and the application of the model described in Figure 5-1 to those post containment failure scenarios may result in a reduction in the risk estimate. The RAI response also states that for "successful pool venting," only sources outside the containment are credited. Clarify how the PRA model described in the response to RAI 5 correctly captures the risk impact from accident scenarios "post containment failure" and those with "successful pool venting."

Enclosure

c. In response to RAI 5.c, .6LERF (large early release frequency) resulting from loss of NPSH was equated to .6CDF. If a new method to estimate .6LERF is deemed necessary in response to items a or b above, describe and justify any credit taken for reducing

.6LERF below the value for .6CDF.

RAI 8-A

Background

RAI 8 requested that EGC explain how it determined that the identified leakage near the containment structure was groundwater and is not impacting the structural integrity or leak-tightness of the containment.

In its response, EGC explained the leakage was not groundwater but did not address how it was determined the leakage was not impacting containment. The response contained no information regarding the leakage location relative to containment, and no discussion of whether or not the leakage is contained within the sand pocket drain lines.

If the leakage is contained within the sand pocket drain lines (i.e., no indications of leakage through concrete in the general vicinity of the drains), the NRC staff understands how it can be determined the leakage is not impacting the containment. However, if leakage is occurring through concrete in the general area of the drains, the leakage could be impacting inaccessible portions of containment. The response did not provide enough information for the NRC staff to determine that the leakage is contained within the drains, or if the leakage is outside the drain lines (as implied in Notes 2 and 4) that the leakage is not impacting containment.

Request For both units, explain how EGC determined that the identified leakage is not impacting the structural integrity or leak-tightness of the containment. The explanation should include whether or not the leakage is contained such that it is not impacting containment and how it was determined that the leakage is not impacting inaccessible portions of the containment (e.g., location of leakage precludes it from contacting containment, minimal amount of leakage or sporadic nature of leakage limits possibility of corrosion).

ML17262A089 NRR-106 OFFICE DORL/LPL3/PM DORL/LPL3/LA DORL/LPL3/BC DORL/LPL3/PM NAME KGreen SRohrer DWrona KGreen DATE 9/19/17 9/19/17 9/19/17 9/19/17