ML15051A409

Response to Request for Additional Information Regarding Permanent Extension of Type a and C Leak Rate Test Frequencies License Amendment Request

ML15051A409
Person / Time
Site:	Calvert Cliffs
Issue date:	02/17/2015
From:	George Gellrich Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML15051A409 (16)

Text

AWExelon Generation, George Gellrich Site Vice President Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657 410 495 5200 Office 717 497 3463 Mobile www.exeloncorp.com george.gellrich@exeloncorp.com 10 CFR 50, Appendix J February 17, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

Subject:

Request for Additional Information Regarding Permanent Extension of Type A and C Leak Rate Test Freauencies License Amendment Reauest

References:

1. Letter from G. H. Gellrich (Exelon Generation) to Document Control Desk (NRC), dated September 18, 2014, License Amendment Request: Revise Technical Specification Section 5.5.16 for Permanent Extension of Type A and C Leak Rate Test Frequencies

2. Letter from N. S. Morgan (NRR) to G. H. Gellrich (Exelon Generation), dated January 22, 2015, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2-Request for Additional Information Regarding Permanent Extension of Type A and C Leak Rate Test Frequencies License Amendment (TAC Nos.

MF4898 and MF4899)

In Reference 1, Exelon Generation submitted a license amendment request for the Calvert Cliffs Nuclear Power Plant to permanently extend the Type A and C leak rate test frequencies. In Reference 2, the Nuclear Regulatory Commission issued a request for additional information.

The responses to the request for additional information are provided in Attachment (1).

Attachment (2) contains a revised marked up copy of the Technical Specification 5.5.16 which replaces the marked up copy provided in Reference 1. Attachment (3) contains a revised risk assessment report that replaces the risk assessment provided in Reference 1.

The responses to the request for additional information contained in this correspondence do not impact the Technical Evaluation conclusions, the No Significant Hazards Consideration, and the Environmental Assessment stated in Reference 1.

There are no regulatory commitments contained in this correspondence.

AýN7 oLL

Document Control Desk February 17, 2015 Page 2 Should you have questions regarding this matter, please contact Mr. Michael J. Fick at (410) 495-6714.

I declare under penalty of perjury that the foregoing is true and correct. Executed on February 17, 2015.

Respectfully, George H. Gellrich Site Vice President GHG/KLG/bjm Attachments: (1) Responses to Request for Additional Information (2) Marked Up Technical Specifications Page (3) Evaluation of Risk Significance of Permanent ILRT Extension cc: NRC Project Manager, Calvert Cliffs NRC Regional Administrator, Region I NRC Resident Inspector, Calvert Cliffs S. Gray, MD-DNR

1 ATTACHMENT (1)

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION Calvert Cliffs Nuclear Power Plant*

February 17, 2015

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION By letter dated September 18, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14265A219), Calvert Cliffs Nuclear Power Plant, LLC submitted a license amendment request (LAR) for Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Calvert Cliffs). The proposed LAR revises Technical Specification 5.5.16, "Containment Leakage Rate Testing Program," to allow an increase in the Type A primary containment integrated leak rate test (ILRT) interval from in the current 10-year frequency to a maximum of 15 years and extension of the Type C containment isolation valve leakage test frequency from 60 to 75 months. The Nuclear Regulatory Commission (NRC) staff is reviewing the application and has determined that the following additional information is needed to complete its review.

ProbabilisticRisk Assessment (PRA) Licensing (APLA) RAI 1:

In the safety evaluation report for Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," the NRC staff, in part, stated that for licensee requests for a permanent extension of the ILRT surveillance interval to 15 years "[clapabilitycategory I of ASME [American Socieity of Mechanical Engineers] RA-Sa-2003 shall be applied as the standard,since approximate values of CDF[core damage frequency] and LERF [large early release frequency] and their distribution among release categoriesare sufficient for use in the EPRI methodology."

Attachment 3 of the LAR states that the 2010 full scope peer review of the internal events PRA model identified three supporting requirements (SRs) that were 'Not Met'. Table 1 of Attachment 3 to the LAR lists the findings from the 2010 peer review, but does not identify the three "Not Met' SRs. Identify which SRs were considered not met. For each SR, summarize why not meeting Capability Category I requirements will have no impact on the ILRT extension application.

CCNPP Response APLA RAI 1:

The three SRs that were noted as "not met" are LE-F2, LE-G5, and IFQU-A10. LE-F2 relates to LERF results. The dominant LERF contributors were reviewed and model changes implemented prior to the ILRT analysis. LE-G5 relates to the documentation of limitations of applications of the PRA. IFQU-A1O relates to documentation of the treatment of the internal flood analysis in the event trees. Therefore, none of the "not met" SRs impact the ILRT extension analysis.

APLA RAI 2:

Section 5.3.2 of Attachment 3 to the LAR uses the Calvert Cliffs methodology from 2002 in evaluating the impact of steel liner corrosion on the extension of ILRT testing intervals. This assessment was based on two observed corrosion events at North Anna Power Station, Unit 2 and Brunswick Steam Electric Plant,Unit 2.

a. If there have been additional instances of liner corrosion that could be relevant to this assessment, provide an updated list of observed corrosion events relevant to Calvert Cliffs containment, and an evaluation of the impact on risk results when all relevant corrosion events are included in the risk assessment.

b. Per EPRI TR-1009325, Revision 2, the risk metrics associated with the ILRT extension applicationinclude changes in LERF, populationdose, and conditional containment failure probability(CCFP). The steel liner corrosionassessment in Section 5.3.2 of Attachment 3 to the LAR calculates only the change in LERF. Include an estimate of change in 1

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION population dose and CCFP due to increase in steel liner corrosion likelihood and demonstrate acceptabilityof the risk results.

CCNPP Response APLA RAI 2:

a. A search of the Licensee Event Report database identified two additional events have occurred since the Calvert Cliffs analysis was performed. In January 2000, a 3/16-inch circular through-liner hole was found at Cook Nuclear Plant Unit 2 caused by a wooden brush handle embedded immediately behind the containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) containment liner caused by pitting originating from the concrete side due to a piece of wood that was left behind during the original construction that came in contact with the steel liner. Two other containment liner through-wall hole events occurred at Turkey Point Units 3 and 4 in October 2010 and November 2006, respectively. However, these events originated from the visible side caused by the failure of the coating system, which was not designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these through-wall liner holes was attributed to corrosion originating from the outside concrete surface due to the presence of rayon fiber foreign material that was left behind during the original construction and was contacting the steel liner. For risk evaluation purposes, these five total corrosion events occurring in 66 operating plants with steel containment liners over a 17.1 year period from September 1996 to October 4, 2013 [i.e., 5/(66*17.1) = 4.43E-03] are bounded by the estimated historical flaw probability based on the two events in the 5.5 year period of the Calvert Cliffs analysis [i.e., 2/(70*5.5) = 5.19E-03] incorporated in the EPRI guidance.

b. An estimate of change in population dose and CCFP due to increase in steel liner corrosion likelihood are provided in the following tables.

Unit 1 Steel Liner Corrosion Sensitivity CCFP CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Baseline 2.04E-01 2.09E-01 2.13E-01 5.11E-03 8.76E-03 3.65E-03 CCFP Corrosion Likelihood 2.04E-01 2.09E-01 2.13E-01 5.16E-03 8.84E-03 3.68E-03 X 1000 Corrosion Likelihood 2.04E-01 2.1OE-01 2.14E-01 5.57E-03 9.56E-03 3.98E-03 x 10000 Corrosion Likelihood 2.06E-01 2.16E-01 2.23E-01 9.76E-03 1.67E-02 1.29E-02 X 100000 2

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION Unit 2 Steel Liner Corrosion Sensitivity CCFP CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-1 0) 1-per-1 5) 1-per-1 5)

Baseline 2.01 E-01 2.06E-01 2.09E-01 5.08E-03 8.71 E-03 3.63E-03 CCFP Corrosion Likelihood 2.01 E-01 2.06E-01 2.09E-01 5.13E-03 8.79E-03 3.66E-03 X 1000 Corrosion Likelihood 2.01 E-01 2.06E-01 2.10E-01 5.54E-03 9.50E-03 3.96E-03 X 10000 Corrosion Likelihood 2.03E-01 2.12E-01 2.19E-01 9.70E-03 1.66E-02 6.93E-03 X 100000 Unit 1 Steel Liner Corrosion Sensitivity Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Dose Rate 3.63E-02 1.21 E-01 1.82E-01 8.48E-02 1.45E-01 6.06E-02 Corrosion Likelihood 3.67E-02 1.27E-01 2.04E-01 9.08E-02 1.67E-01 7.61 E-02 X 1000 Corrosion Likelihood 3.96E-02 1.84E-01 4.01E-01 1.45E-01 3.61E-01 2.17E-01 X 10000 Corrosion Likelihood 6.94E-02 7.53E-01 2.37E+00 6.84E-01 2.30E+00 1.62E+00 X 100000 Unit 2 Steel Liner Corrosion Sensitivity Dose Rate CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Dose Rate 1.96E-02 6.55E-02 9.82E-02 4.58E-02 7.86E-02 3.27E-02 Corrosion Likelihood 1.98E-02 6.89E-02 1.10E-01 4.91E-02 9.02E-02 4.12E-02 X 1000 Corrosion Likelihood 2.14E-02 9.96E-02 2.17E-01 7.82E-02 1.95E-01 1.17E-01 X 10000 Corrosion Likelihood 3.75E-02 4.07E-01 1.28E+00 3.70E-01 1.25E+00 8.76E-01 X 100000 APLA RAI 3:

Section 4.2.7 of EPRI TR-1009325, Revision 2-A states that "[w]here possible, the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended ILRT intervals." The EPRI TR-1009325, Revision 2-A further states that the "assessment can be taken from existing, 3

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION previously submitted and approvedanalyses or another alternatemethod of assessingan order of magnitude estimate for contribution of the external event to the impact of the changed interval." Section 5.3. 1 in Attachment 3 to the LAR assesses the potential impact from external events contribution.

a. The results of the seismic PRA performed for the Individual Plant Examinations for External Events (IPEEE) were used to assess the seismic risk with a reported CDF of 1.07E-5/year for both units. Section 8, "Summary and Conclusions," of the IPEEE report (CalculationNo. RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "IndividualPlant Examination of External Events," August 1997) reports seismic CDF values of 1.29E-5/year for Calvert Cliffs 1 and 1.52E-5/year for Calvert Cliffs 2. Justify the use of the 1.0 7E-5/year seismic CDF value in the external events sensitivity study.

b. In Section 5.3.1.1 of Attachment 3 to the LAR, the results from the IPEEE fire analysis were used to assess fire risk (CDF of 1. COE-5/year and LERF of 7.15E-7/year for Calvert Cliffs 2). Section 8, "Summary and Conclusions," of the IPEEE report (Calculation No.

RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "Individual Plant Examination of External Events," August 1997) reports a fire CDF of 9.6E-5/year for Calvert Cliffs 2.

Justify the use of selected IPEEE Calvert Cliffs 2 fire CDF/LERF values and discuss acceptability of Calvert Cliffs 2 risk results when using the IPEEE fire CDF/LERF.

CCNPP Response APLA RAI 3:

a. Data from Table 3-6 of the IPEEE Seismic Analysis is used to calculate a Class 3b frequency due to seismic. As noted in Table 3-6 of the IPEEE Seismic Analysis, the values given in Table 3-6 reflect quantification without the surrogate top event LA. Top event LA represents seismic failure of rugged plant systems at a conservative screening fragility. Therefore, the total CDF is higher than the 1.07E-05/yr value given in Table 3-6. The CDF values given in Section 8.1 of the IPEEE Seismic Analysis are 1.29E-5/yr for Unit 1 and 1.52E-5/yr for Unit 2.

The CDF contribution from surrogate top event LA was not included in the Unit 1 containment failure frequencies provided in the IPEEE (no containment failure frequencies are provided for Unit 2). In lieu of justification for the value of 1.07E-5/year seismic CDF, the contribution is conservatively added to the Unit 1 CDF of 1.07E-5/yr from Table 3-6 of the IPEEE to Containment Category I Failure (Intact). Note that the Intact category CDF is slightly rounded so that the total seismic CDF is preserved. Then, the percent each category contributes to the total CDF is calculated for the Unit 1 values and applied to the Unit 2 values because it is assumed that Unit 2 would have similar containment failure fractions to Unit 1. The resulting containment failure frequencies and total external events contribution using the revised IPEEE seismic CDF for each Unit are shown in the following tables:

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION Seismic Contribution to Frequencies of Containment Failure Categories Containment Failure Unit 1 Seismic CDF Percent of CDF Unit 2 Seismic CDF (/yr)

Category (/yr)

I. Intact Containment 2.69E-06 20.85% 3.17E-06 I1.Late Containment Failure 8.63E-06 66.90% 1.02E-05 111.Early Small Containment 1.70E-07 to 1.27E-06 1.32% to 9.84% 2.00E-07 to 1.50E-06 Failure IV. Early Large Containment 3.13E-07 to 1.41 E-06 2.43% to 10.93% 3.69E-07 to 1.66E-06 Failure V. Small Containment Bypass 0 0% 0 VI. Large Containment Bypass 0 0% 0 Total 1.29E-05 1.52E-05 CCNPP Unit 1 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year 1 per 10 year 1 per 15 years per 15 years)

External Events 5.13E-08 1.71 E-07 2.57E-07 2.05E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 6.27E-08 2.09E-07 3.13E-07 2.51 E-07 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year 1 per 10 year 1 per 15 years per 15 years)

External Events 6.52E-08 2.17E-07 3.26E-07 2.61 E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 7.13E-08 2.38E-07 3.57E-07 2.85E-07 5

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

b. In lieu of justification of selected fire IPEEE CDF and LERF, the risk results have been revised using a Unit 2 fire CDF of 9.6E-5/year as given in Section 8 of the IPEEE. The Unit 2 containment failure frequencies and frequency 3b change are shown in the following tables:

Unit 2 Fire Contribution to Frequencies of Containment Failure Categories Containment Failure Category Percentage Unit 2 Fire CDF (/yr)

I. Intact Containment 36.4% 3.50E-05 II. Late Containment Failure 55.5% 5.33E-05 Ill. Early Small Containment Failure 1.7% 1.58E-06 IV. Early Large Containment Failure 6.5% 6.13E-06 V. Small Containment Bypass 0.0% 0.00E+00 VI. Large Containment Bypass 0.0% 0.00E+00 Total 9.60E-05 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year 1 per 10 year 1 per 15 years per 15 years)

External Events 2.40E-07 8.00E-07 1.20E-06 9.61 E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 2.46E-07 8.21 E-07 1.23E-06 9.88E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.0E-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events PRA, the total LERF values are calculated below:

Unit 2: LERFU2 = LERFU2internal + LERFU2seismic + LERFU2fire+ LERFU2class3Bincrease

= 1.56E-6/yr + 1.66E-6/yr + 6.13E-6/yr + 9.84E-7/yr = 1.03E-5/yr The Unit 2 LERF is barely greater than 1.OE-5. However, the Unit 2 Seismic LERF is between 3.69E-07 and 1.66E-06, and the highest Seismic LERF value was conservatively used to calculate 1.03E-5. If the 74th percentile or smaller value of this range (< 1.32E-6) is used, the total Unit 2 LERF is less than 1.OE-5. Moreover, the IPEEE does not include recent significant plant modifications designed specifically to reduce fire risk. Therefore, it is reasonable to conclude that the total Unit 2 LERF is less than 1.OE-5. Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION APLA RAI 4:

Section 5.2.4 of Attachment 3 to the LAR refines the calculation of the Class 3b frequencies for internal events by examining the source term. The conservatism in Class 3b frequency is reduced by analyzing the source term release time and defines an early release as occurring before 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which allows the removal of three accident scenarios from the Class 3b frequency. Elimination of these three scenarios appears to reduce the Class 3b frequency for internal events by a factor of 3 to 5. Section 5.3.1 of Attachment 3 to the LAR indicates that the same approach is used in the calculation of the fire Class 3b frequency in the external events sensitivity study when using the NFPA-805 fire PRA.

a. Provide the calculated timing of the expected release for each of these three accident scenarios.

b. Provide the basis for the 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delineationbetween early and late release.

c. Explain whether releases from these scenarios were included in the analysis to calculate the increase in the total integrateddose risk for all accident sequences.

CCNPP Response APLA RAI 4:

a. The calculated timing of the expected release for the scenarios is as follows:

HRIF: 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> GIOY: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> MRIF: 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />

b. The 6.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delineation between early and late release is based on the calculation of evacuation time estimates for Calvert Cliffs Nuclear Power Plant. Since early release timing is defined by time short enough that ability to evacuate nearby population is impaired such that a fatality is possible, and the calculation shows that 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is sufficient for evacuation, this is chosen as the delineation between early and late release.

c. The releases from these scenarios were excluded in the analysis to calculate the total integrated dose risk, since the scenarios are excluded from the class 3b contribution based on either containment spray success or the late timing of the release.

Mechanicaland Civil Engiineering(EMCB) RAI 1:

Section 3.8 of the LAR states that, (1) "at Calvert Cliffs, a test pipe was provided for each continuous segment of the bottom liner plate weld chase test channels;"and (2) an analysis of the NRC Information Notice 2014-07, "Degradationof Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" "is in progress to determine if this issue is applicable to CalvertCliffs."

Discuss (1) the results of the Calvert Cliffs evaluation relative to the NRC Information Notice 2014-07 and (2) the operating experience, inspection results, and any corrective actions relative to the bottom floor liner plate weld leak chase test channels, including the existing test pipe.

CCNPP Response EMCB RAI 1:

The Calvert Cliffs design configuration is different from the one considered as "typical" in information notice IN 2014-07. At Calvert Cliffs, the pipe protruding from the leakchase channel 7

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION is either "flush" to the floor or exposed to the surface with no access box and cover plate. This configuration eliminates the chances of moisture accumulation thus avoiding degradation of this area.

The Containment Inservice Inspection Program includes the ASME Code, Section Xl, Subsection IWE inspection requirements for containment liner as mandated by 10 CFR 50.55a, "Codes and Standards." The inspection frequency required by ASME Section Xl Subsection IWE is to perform a General Visual inspection of all accessible areas in Containment, including the moisture barrier, each inspection period, which may consist of more than 1 refueling outage (RFO). At Calvert Cliffs the inspection is performed every RFO, which meets and exceeds the ASME Code inspection frequency requirements in the instances where the period contains more than 1 RFO. The examination includes the inspection of the leak chase channels caps/plug and exposed portions of pipes during the General Visual inspection of 10' containment liner and 10' moisture barrier.

In addition the Containment Leakage Rate Testing Program requires removal of the leak-chase channel pipe cap/plug for functional testing of containment during the local leak rate test (LLRT)/ILRT per procedures STP-M-471-1/2 and STP-M-662-1/2, respectively. A satisfactory LLRT/ILRT exam indicates containment leak tightness and leak-chase channel system integrity.

The procedures include steps for pipe caps/plugs removal and installation.

EMCB RAI 2:

Please confirm the following:

a. There are no planned modifications for Calvert Cliffs that will require a Type A test prior to the next units Type A test proposed in this LAR.

b. There is no anticipated addition or removal of plant hardware within the containment building which could affect its leak-tightness.

CCNPP Response EMCB RAI 2:

There are no planned modifications for Calvert Cliffs Units 1 and 2 that will require a Type A test prior to the next Unit 1 and 2 Type A test proposed under this LAR, and there is no anticipated addition or removal of plant hardware within the containment building, which could affect its leak-tightness.

EMCB RAI 3:

Table 3.4.5 of the LAR indicates that the Calvert Cliffs 2 containment moisture barrier seal requires an augmented examination. Section 3.5 "OperatingExperience" of the LAR indicates that the augmented examination of moisture barrierseal is due to a crack identified during the 2013 refueling outage.

Discuss the extent of the degradation of the moisture barrier seal and provide information relative to the inspection of the liner area at the wall to floor transition under the moisture barrier seal where the 2013 crack was identified.

CCNPP Response EMCB RAI 3:

A 1" crack was identified on the moisture barrier near leak chase channel no. 8. It was determined that there was no liner (carbon steel) impact due to moisture or evidence of 8

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION moisture intrusion. The crack was repaired under work order C921941 10 and a subsequent examination is scheduled for performance during the 2015 Unit 2 RFO.

EMCB RAI 4:

Section 3.5.2 "Containment Concrete" of the LAR indicates that during the 2005 and 2007 concrete inspections, new tendon grease leakage was identified. Discuss Calvert Cliffs root cause evaluation and corrective actions taken to disposition this inspection finding.

CCNPP Response EMCB RAI 4:

The evaluation of the identified tendon grease leakage categorized the findings as of 'low safety significance' and characterized the leakage as grease cap leaks requiring cleaning and/or joint sealant replacement. A root cause evaluation was not required. All containment concrete related degradation issues (including grease leaks on containment dome) were addressed during the 2014 Containment Concrete Repair Project at Calvert Cliffs. The concrete and grease cap sealant joint repairs were performed and completed under work orders C91864927 and C91864929.

EMCB RAI 5:

Pleaseprovide the following information:

a) Percentof the total number of Type 8 tested components that are on 120-month extended performance-basedtest interval.

b) Percent of the total number of Type C tested components that are on 60-month extended performance-basedtest interval.

CCNPP Response EMCB RAI 5:

a) Percent of the total number of Type B tested components that are on 120-month extended performance-based test interval is 67 of 77 for Unit 1 and 65 of 77 for Unit 2 for a station total of 86%.

b) Percent of the total number of Type C tested components that are on 60-month extended performance-based test interval is 53 of 73 for Unit 1 and 34 of 73 for Unit 2 for a station total of 60%.

Containmentand Ventilation RAI 1:

Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50, Appendix J, Option B Implementation item #3 states that:

The regulatory guide or other implementation document used by a licensee or applicant for an operatinglicense under this part or a combined license under part 52 of this chapter to develop a performance-based,leakage-testing program must be included, by general reference, in the plant technicalspecifications.

In addition, in the August 20, 2013, letter from the NRC to Nuclear Energy Institute (NEI)

(ADAMS Accession No. ML13192A394), the NRC staff indicated that.

Due to the omission of the limitations and conditions from NEI 94-01, Revision 2 SE into NEI 94-01, Revision 3-A, the NRC will not be able to reference NEI 94-01, Revision 3-A in the update to RG" 1.163. Any licensee submissions referencing the TR will require 9

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RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION requests for additionalinformation from the NRC to address the limitations and conditions from the NRC SE for NEI 94-01, Revision 2.

Section 2.0, "DetailedDescription,"of the LAR included a reference to NEI 94-0 1, Revision 3-A.

However, the conditions and limitations contained in NEI 94-01, Revision 2-A, were not incorporatedin NEI 94-01, Revision 3-A.

Explain how the 10 CFR 50 Appendix J, Option B Implementation item #3 and NRC concerns stated in the August 20, 2013, letter from the NRC to NEI will be addressed.

CCNPP Response Containment and Ventilation RAI 1:

The referenced letter stated the following:

The NRC based its conclusion, in part, on the NEI 94-01, Revision 3-A TR Executive Summary which notes that the TR meets the limitations and conditions of the SE for both Revision 2 and Revision 3. Revision 2-A of NEI 94-01 was issued in 2008 and included provisions for extending the Integrated Leak Rate Testing (ILRT) interval to 15 years subject to the limitations and conditions provided in the SE for Revision 2. Revision 3-A was issued in July 2012 and included guidance for extending the Type C Local Leak Rate Testing (LLRT) interval to 75 months. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

In Revision 2-A, the Executive Summary states:

This document, NEI 94-01, describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER is included in the front matter of this report.

That statement was deleted in Revision 3-A, and replaced with the following:

This document, NEI-94-01, Revision 3-A describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to fifteen years. NEI 94-01 has been endorsed by Regulatory Guide 1.163 (September 1995) and NRC Safety Evaluations of June 25, 2008 and June 8, 2012 as an acceptable methodology for complying with the provisions of Option B to 10 CFR Part 50. The regulatory positions stated in Regulatory Guide 1.163 (September 1995) as modified by NRC Safety Evaluations of June 25, 2008 and June 8, 2012 are incorporated in this document.

In response to the above the following change is proposed.

Section 2.0 of Reference 1 stated the following when submitted:

The proposed change to Calvert Cliffs TS 5.5.16, "Containment Leakage Rate Testing Program" will remove exceptions (a), (b), and (c) and replace the reference to RG 1.163 with a reference to NEI Topical Report NEI 94-01 Revision 3-A. The proposed change will revise TS 5.5.16 to state, in part:

"A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall 10

ATTACHMENT (1)

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing 'Performance-Based Option of 10 CFR Part 50, Appendix J,"

dated July 2012."

To address the RAI, the proposed change to Section 2.0 of Reference 1 should be worded as follows:

The proposed change to Calvert Cliffs TS 5.5.16, "Containment Leakage Rate Testing Program" will remove exceptions (a), (b), and (c) and replace the reference to RG 1.163 with a reference to NEI Topical Report NEI 94-01 Revision 3-A. The proposed change will revise TS 5.5.16 to state, in part:

"A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008."

A marked up copy of TS 5.5.16 is contained in Attachment (2). This marked up copy replaces the marked up copy originally submitted in Reference 1.

11

ATTACHMENT (2)

MARKED UP TECHNICAL SPECIFICATIONS PAGE Calvert Cliffs Nuclear Power Plant February 17, 2015

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained iriWIRe-a~uloy .... d 1.163, "Perf,,-ua -e 4neluding crr-ata, as modificd by the follown - cptOnz:

~.Nuelear Enrgry institu.te (NEI) 94 01 1995, Sectioni 9.2.3.

The first Unit 1 Type A test performed aftcr the June 15, t992 Tye A-test shall be prformed no latzr than Junr 14-,

2007. The first Unit 2 Type A test performed after the

b. Unit 1 is excefted from post modification integrated leakage-Pa~te testing reuiemnts associateed with steam gener-ator

-replee07t-.

ce. Unit 2 is emeepted frem post modification integrated leakage rate testing requirements associatcd with steam gcnerato r~ep! eeementrv The peak calculated containment internal pressure for the design basis loss-of-coolant accident, P,, is 49.7 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, L., shall be 0.16 percent of containment air weight per day at P,.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is _<1.0 L,.

During the first unit startup following testing, in accordance with this program, the leakage rate acceptance T-v!,,R r I criterion are _*0.60 L. for Types B and C tests and <_0.75 L, for Type A tests. , --- --

-/0 -,i 4 Pnr 6 A -[ i Op ,.'P CALVERT CLIFFS - UNIT 1 5.5-18 Amendment No. 308 CALVERT CLIFFS - UNIT 2 Amendment No. 286