Text

Response to NRC Clarification Request for NFPA-805 License Amendment Request
	ML16117A499
Person / Time
Site:	Calvert Cliffs
Issue date:	04/22/2016
From:	George Gellrich; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML16117A499 (8)
	v • d • e

SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10 CFR 2.390 April 22, 2016 George H. 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 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Calvert Cliffs Nuclear Power Plant, Units No. 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRG Docket Nos. 50-317 and 50-318 Response to NRG Clarification Request for NFPA-805 License Amendment Request

Reference:

1.

Letter from G.H Gellrich (CCNPP) to Document control Desk (NRG), dated September 24, 2013, License Amendment request re: Transition to 1 O CFR 50.48(c) - NPPA 805 Performance Based Standard for Fire Protection In Reference 1, Calvert Cliffs Nuclear Power Plant submitted a license amendment request to adopt National Fire Protection Association 805. On April 7, 2016, a conference call was held with the Nuclear Regulatory Commission staff to discuss some clarifications requested by the Nuclear Regulatory Commission staff.

Calvert Cliffs' responses, as discussed during the conference call, are contained in Attachment (1). Enclosure 1 contains page replacements for Attachments G and W of the license amendment package. also contains Attachment S with all revisions incorporated and updated to show items completed.

Attachments S and W in Enclosure 1 contain security-rnlated information and are requested to be withheld from public disclosure under 1 O CFR 2.390, This additional information does not change the No Significant Hazards Determination provided in Reference 1. No regulatory commitments are contained in this letter.

Should you have questions regarding this matter, please contact Mr. Larry D. Smith at (410) 495-5219.

Upon removal of Attachment S and W pages in Enclosure 1, this submittal is not restricted

Document Control Desk April 22, 2016 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on April 22, 2016.

Respectfully,

~~-~

George H. Gellrich Site Vice President GHG/KLG/bjm

Attachment:

(1) Responses to NRC Request for Clarification cc:

Enclosure:

1 Page Replacements for Attachment G, S, and W (Without Enclosures)

NRC Project Manager, Calvert Cliffs NRC Regional Administrator, Region I NRC Resident Inspector, Calvert Cliffs S. Gray, MD-DNR

ATTACHMENT {1)

RESPONSES TO NRC REQUEST FOR CLARIFICATION Calvert Cliffs Nuclear Power Plant April 22, 2016

ATTACHMENT (1)

RESPONSES TO NRC REQUEST FOR CLARIFICATIONS

1.

Clarification to PRA 03.b in reference to PRA RAI 02.b.11.01 In the response to PRA RAI 03.b, dated February 24, 2016, the disposition of PRA RAI 02.b.ii.01 (page 31 in Attachment 1 of the response) indicates that the final number of joint HEPs less than 1.0E-5 and several example justifications that were requested by PRA RAI 02.b.ii.01 but not provided in the July 6, 2015 response would be provided in a supplemental response to PRA RAI 02.b.ii.01 or as part of the response to PRA RAI 03. However, the NRC staff cannot find this information.

Calvert Cliffs Response:

No joint HEP values of less than 1 E-05 were used in the final quantification, therefore, no supplemental response was provided.

2.

Clarification to PRA RAI 19.01 PRA RAI 19.a.i noted several inconsistencies associated with LAR Table W-7 in which some deterministically compliant fire areas (i.e., NFPA Basis 4.2.3.2) were indicated as having VFDRs and some performance-based fire areas (i.e., NFPA Basis 4.2.4.2) were shown to have neither VFDRs nor a calculated delta risk (i.e., N/A). Although the response to PRA RAI 19, dated February 24, 2016, indicates that such inconsistencies have been resolved, they still appear to be present in the update provided to Table W-7. For instance, Fire Area 2 is noted as being deterministically compliant (i/e., NFPA Basis 4.2.3.2); however, a delta risk is calculated in Table W-7, and VFDRs are identified in the updated Attachment C. The reverse is true for Fire Areas 3 and 4. The NRC staff seeks clarification if these are simply editorial errors or if they could impact the reported estimates.

Calvert Cliffs Response:

These are editorial errors in the column in question. In Table W-7, Fire Areas with VFDR =

"Yes" are performance based (Section 4.2.4.2), while fire areas with VFDR = "No" are deterministically compliant (4.2.3.2).

Where VFDR = "Yes" appears, there are delta risk numbers.

Where VFDR ="No" appears, the delta risk is "N/A." A revised Table W-7 is provided. Also one editorial change is made to Table W-6 for fire area AB-1 to note that there is no CDF or LEAF number for this fire area. This change is also attached.

3.

Clarification regarding VFDR 16-27-1 (Fire Area 16) and VFDRs 16-61-2 and 17-25-2 (Fire Area 17)

In the recent revisions provided in Attachment C, the licensee documented that risk, DID and SM criteria were met with no further action related to VFDR 16-27-1, (ADVs 1 CV-3938 and 1CF-3939) (for Fire Area 16), and VFDRs 16-61-2 and 17-25-2 (ADVs 2CV-3938 and 2CV-3939) (for Fire Area 17) citing the use of the Safety Relief Valves for controlling Steam Generator pressure. However, the revisions provided in Attachment G create inconsistencies with the-approach used to disposition the VFDR.

Consistent with the VFDR disposition in Attachment C, in LAR Attachment G, Table G-1, the licensee removed the recovery actions for fire area 16 to reposition 11 and 12 ADV hand valves in order to locally operate ADVs 1 CV3938 an 1 CV3939 from Panel 1 C43, but inconsistent with the disposition in Attachment C, maintained the PCS action to initialize the ADV controls at panel 1 C43 and use the ADV hand controllers to control RCS temperature and verify natural circulation. In fire area 16, the licensee also removed the action to close the 21 and 22 steam generator ADVs 2CV3939 and 2CV3938, respectively, from panel 2C43.

Similarly, for fire 1

ATTACHMENT (1)

RESPONSES TO NRC REQUEST FOR CLARIFICATIONS area 17, the licensee removed the recovery actions to reposition 21 and 22 ADV hand valves in order to locally operate the ADVs 2CV3938 and 2CV3939 from panel 2C43, but maintained the PCS action to initialize the ADV controls at panel 2C43 and to use the ADV hand controllers to control RCS temperature and verify natural circulation.

In fire area 17, the licensee also removed the-action to close the 11 and 12 steam generator ADVs 1 CV3939 and 1 CV3938, respectively, from panel 1 C43.

However, in the summary of results for Step 1 in the LAR Attachment G, the licensee still states that the hand valves associated with the ADV hand controller on panels 1 C43 and 2C43, respectively, are required to be repositioned in order to enable operation of the ADVs from panels 1 C43 and 2C43.

The NRG staff seeks clarification (1) whether the hand controller at the PCS panels 1 C43 and 2C43 for the steam generator ADVs for both units are credited to meet the nuclear safety performance criteria for a fire in fire area 16 or 17, and (2) whether the description of the PCS actions in the results of Step 1, which are associated with the steam generator ADVs, is accurate.

Calvert Cliffs Response:

The hand controller at the PCS panels 1 C43 and 2C43 for the steam generator ADVs for both units are NOT credited to meet the nuclear safety performance criteria for a fire in fire area 16 or

17. The description of the local actions required to enable the 1 C43 and 2C43 hand controllers on pages G-4 and G-7 is accurate but these actions are not a (NFPA 805) recovery action, since these actions are not required to reduce risk or for defense in depth in fire area 16 or 17.

Attachment C and Attachment G, Table G-1 are correct.

Revised text for the main body of Attachment G is provided.

2 Page Replacements for Attachment G, S, and W Calvert Cliffs Nuclear Power Plant April 22, 2016

Constellation Energy Nuclear Group Attachment G - Recovery Action Transition 33.11 Reactor Coolant Pressurizer Level Indication (1LI11 OX) 34.11 Reactor Coolant Pressurizer Level Indication (1 Ll11 OY)

35. 11 Reactor Coolant Pressurizer Pressure Indication (1Pl105AA) 36.11 Reactor Coolant Pressurizer Pressure Indication (1Pl105B)

37. Unit 1 Neutron Power, Logarithmic, Wide Range, % Power Indication (1 NI016 from either 1 NE002 or 1 NE004 via Hand Switch 1 HS015B)

38. Unit 1 Neutron Power, Logarithmic, Wide Range, Counts Per Second Indication (1NI015 from either 1NE002or1NE004 via Hand Switch 1HS015B)

39. Backup Pressurizer Heater Bank 11 (1 UCC2) Transfer I Control Hand Switch

( 1 HS 100-4A) see note 1 40.Backup Pressurizer Heater Bank 13 (1UH1) Transfer I Control Hand Switch (1HS100-6A) see note 1 41. Unit 1 Reactor Coolant Pump Controlled Bleedoff Isolation Valve 1 CV505 Hand Switch (1 HS2505A)

42. Unit 1 Reactor Coolant Letdown Isolation Valve 1 CV516 Hand Switch (1 HS2516A)

43. Unit 1 Reactor Coolant Sampling Isolation Valve 1 CV4564 Hand Switch (1 HS5464B) 1C43 Notes:

Note 1: Enabling of each Backup Pressurizer Heater also requires a local recovery action to verify closed I reclose the associated feeder breaker to the I UC heater Motor Control Center (1MCC109PH / 1 MCC111 PH) at the 480V Unit Bus (1 BUS1 B01 B / 1 BUS1 B04B) as identified below:

Backup Heater Bank 11 (1 UCC2)

Bank 13 (1UH1)

Feed Breaker 1BKR52-1127 1BKR52-1427 Feed Breaker(s)

Room Location 317 430 Note 2: To enable the 1 C43 hand controller requires a local reso 1.iery action to I UC reposition the associated hand valve(s) as identified below:

Hand Controller Hand Valve(s)

Hand Valve(s)

Room Location 1 HC4056A 1HC4056B 1 HC4511B 1HC4512B 1 HVMS-3938A 1 HVMS-39388 1 HVMS-3939A 1 HVMS-39398 1 HVAFW-4511 1HVAFW-4512 430 430 430 430 226 226 G-4

Constellation Energy Nuclear Group Attachment G - Recovery Action Transition

43. Unit 2 Reactor Coolant Sampling Isolation Valve 2CV4564 Hand Switch (2HS5464B) 2C43 Notes:

Note 1: Enabling of each Backup Pressurizer Heater also requires a local recovery action to verify closed I reclose the associated feeder breaker to the I UC heater Motor Control Center (2MCC209PH / 2MCC211 PH) at the 480V Unit Bus (2BUS2B01 BI 2BUS2B04B) as identified below:

Backup Heater Bank 21 (2UCC2)

Bank 23 (2UH1)

Feed Breaker 2BKR52-2127 2BKR52-2427 Feed Breaker(s)

Room Location 311 407 Note 2: To enable the 2C43 hand controller requires a local recovery action to I uc reposition the associated hand valve(s) as identified below:

Hand Controller Hand Valve(s)

Hand Valve(s)

Room Location 2HC4056A 2HVMS-3939A 407 2HVMS-39398 407 2HC4056B 2HVMS-3938A 407 2HVMS-3938B 407 2HC45118 2HVAFW-4511 205 2HC45128 2HVAFW-4512 205 2HC4525B 2HVAFW-4525 205 2HC45358 2HVAFW-4535 205 2HC3987B 2HVMS-3987 605 2HC39898 2HVMS-3989 605 Note 3: Enabling of Channel A WRNI at 2C43 requires placing 2HS001A1 to off.

This is considered a PCS action that initiates control of instrumentation at the alternate shutdown panel. The hand switch is located in the same room as 2C43.

1 C43 (Unit 1) and 2C43 (Unit 2) are the Primary Control Station for implementation of the Alternate Shutdown Strategy in the event of a fire that requires the evacuation of the Main Control Room. NRC approval for the design of the Alternate Shutdown Panel(s),

and for the overall Alternate Shutdown Strategy to meet the requirements of 10 CFR 50 Appendix R, Section 111.G.3, was provided in SER Supplement No. 3, dated September 27, 1982. Baltimore Gas and Electric Company, Calvert Cliffs Units 1 and 2, Docket Nos., STN 50-317 and 50-318, Item 3.2.1 of the Fire Protection Safety Evaluation Report, Appendix R to 10 CFR Part 50, Items 111.G.3 and 111.L.

G-7