Text

Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request
	ML17159A847
Person / Time
Site:	Farley
Issue date:	06/08/2017
From:	Wheat J; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-17-1067
	Download: ML17159A847 (27)
	v • d • e

A Southern Nuclear Justin T. Wheat Nuclear Licensing Manager 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 205 992 5998 tel 205 992 7601 fax jtwheat @southemco.com June 8, 2017 Docket Nos.: 50-348 NL-17-1067 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Units 1 and 2 Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request Ladies and Gentlemen:

By letter dated November 22, 2016 (Accession Number ML16336A024), Southern Nuclear Operating Company (SNC) requested Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of Joseph M. Farley Nuclear Plant that support a full scope application of an Alternative Source Term (AST) methodology. Proposed Technical Specification changes, which are supported by the AST Design Basis Accident radiological consequence analyses, were included in the license amendment request. In addition, the proposed amendment incorporated Technical Specification Task Force (TSTF)

Traveler, TSTF-448-A, "Control Room Habitability," Revision 3, and TSTF-312-A, "Administrative Control of Containment Penetrations," Revision 1.

By letter dated March 24, 2017, the NRC informed SNC that the NRC has determined additional information is needed to complete its review. On May 23 2017, SNC submitted answers to 39 of the 42 Requests for Information (RAI).

Enclosed are SNC's remaining responses to the RAis (numbers 15, 37, and 40).

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

U. S. Nuclear Regulatory Commission NL-17-1 067 Page2 Mr. Justin T. Wheat states that he is the Nuclear Licensing Manager of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted,

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~.Wheat Nuclear Licensing Manager JTW/NDJ/LC Sworn to nd subscribed before me this _j_ day of 'tf~ '2017.

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My commission expires: ....:;1_,6:......_

Enclosures:

1. SNC Response to Request for Additional Information {# 15, 37 & 40)

2. FNP Unit 1 and Unit 2 Technical Specifications Clean-Typed Pages

3. FNP Unit 1 and Unit 2 Technical Specifications Marked-Up Pages cc: Regional Administrator, Region II NRR Project Manager- Farley Senior Resident Inspector- Farley RTYPE: CFA04.054

Joseph M. Farley Nuclear Plant Units 1 and 2 Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request Enclosure 1 SNC Response to Request for Additional Information(# 15, 37, 40) to NL-17-1067 SNC Response to Request for Additional Information RAI No. 15 {FHA)

Regulatory Basis numbered 1, 2, 5, and 8 apply to RAI No. 15.

RG 1.183 Appendix B regulatory position 5.3 states:

If the containment is open during fuel handling operations {e.g., personnel air lock or equipment hatch is open), 3 the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

3 The staff will generally require that technical specifications allowing such operations include administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation.

In the LAR, Table C of Enclosure 5 states FNP's conformance with RG 1.183 Appendix B.

Table C states that FN P's analysis for RG 1.183 regulatory position 5.3, "Conforms - The FHA radiological release is over a two-hour period." However, the NRC staff could not find two evaluations for the FHA occurring in containment. The first evaluation missing is an evaluation of the FHA in containment with the PAL open and the equipment hatch closed. The second evaluation missing is an evaluation of the FHA in containment with the equipment hatch open and the PAL closed.

FNP TS 3.9.3 allows three different configurations during core alterations and during movement of irradiated fuel assemblies within containment. The configurations are: {1) equipment hatch open and capable of being closed and held in place by four bolts, and PAL with one or more door{s) closed, {2) equipment hatch closed and held in place by four bolts, and PAL open with one door in the air lock capable of being closed, and {3) equipment hatch open and capable of being closed and held in place by four bolts and PAL open with one door in air lock capable of being closed. SNC submitted a FHA in containment analysis that addresses the configuration of the equipment hatch open and the PAL open.

Consistent with the allowances of TS 3.9.3 please provide:

1. An evaluation that is consistent with RG 1.183 and meets the limits in RG 1.183, SRP 15.0.1 and 10 CFR 50.67 and evaluates the FHA in containment with the equipment hatch open and the PAL closed.

2. An evaluation that is consistent with RG 1.183 and meets the limits in RG 1.183, SRP 15.0.1 and 10 CFR 50.67 and evaluates the FHA in containment with the equipment hatch closed and the PAL open.

If it is SNC's intent to be able to move irradiated fuel assemblies within the containment with onlvthe configuration of equipment hatch open and PAL open, please discuss TS 3.9.3 and limits to other configurations.

E1-1 to NL-17-1067 SNC Response to Request for Additional Information SNC Response to RAI No. 15:

SNC has analyzed the FHA in containment accident for the following three cases: Containment Equipment Hatch (EH) open with the Personnel Airlock (PAL) open, EH open with the PAL closed, and EH closed with the PAL open. Because of the inclusion of the two additional cases, and consideration of the comments in the technical basis, above, SNC must revise Enclosure 1 table 3.6a (pages E1-1 0 and E1-11) and Enclosure 7 of the LAR submittal. Both revisions are shown below:

Table 3.6a, pages E1-10 & E1-11 Reactor Power 2,831 MWt Fraction of Fission Product Inventory in Gap 1-131 0.08 Kr -85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals (retained in pool) 0.12 Number of Damaged Fuel Assembly 1 Irradiated Fuel Decay 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Radial Peaking Factor 1.7 Iodine Chemical Forms Released from Fuel to pool Elemental 99.85%

Organic 0.15%

Minimum Refueling Cavity and Pool Water Depths 23 feet Decontamination Factors Noble Gases 1 Particulates Infinity Overall Iodine 200 Iodine Chemical Forms Released from Pool Elemental 57%

Organic 43%

Duration of Release 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Atmospheric Dispersion Factors (sec/m 3 )

Containment Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 8.79E-04 2-8 1.10E-4 6.77E-04 8-24 3.32E-04 Plant Vent Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 1.62E-03 2-8 1.10E-4 1.37E-03 8-24 7.10E-04 E1-2 to NL-17-1067 SNC Response to Request for Additional Information Open Equipment Hatch & Open Personnel Airlock Case Off Site Doses (EAB and LPZ) Model Containment Hatch Flow Rate 55,000 cfm Personnel Airlock Flow Rate (see note below) 25,000 cfm Control Room Ingress/Egress Dose Model Containment Hatch Flow Rate 43,500 cfm Auxiliary Building Mixing Volume 100,650 cubic feet Personnel Airlock Flow Rate (see note below) 12,300 cfm Control Room Containment Hatch Dose Model Containment Hatch Flow Rate 55,000 cfm Personnel Airlock Flow Rate 12,300 cfm Control Room Personnel Airlock Dose Model Containment Hatch Flow Rate 43,500 cfm Personnel Airlock Flow Rate (see note below) 25,000 cfm Open Equipment Hatch & Closed Personnel Airlock Case Containment Hatch Flow Rate 55,000 cfm Closed Equipment Hatch & Open Personnel Airlock Case Control Room Ingress/Egress Dose Model Auxiliary Building Mixing Volume 100,650 cubic feet Personnel Airlock Flow Rate (see note below) 12,300 cfm Control Room Personnel Airlock Dose Model Personnel Airlock Flow Rate (see note below) 25,000 cfm Note about Personnel Airlock flow rates: For the three model elements where release to the environment should be conservatively maximized, the flow rate through the Personnel Airlock is assumed to be 25,000 cfm and passes directly to the environment through the Plant Vent Stack as drawn by the normal auxiliary building HVAC system. This flow rate assumption conservatively increases the release to the environment and back into the CR through the CREFS intakes. For the model elements where the ingress/egress through the CR doors is modelled, the flow out the airlock into the auxiliary building hall is 12,300 cfm, which is based on sensitivity studies and the capacity of the normal HVAC system. The studies show that this flow rate maximizes the auxiliary building radioactivity concentrations, and causes the highest dose from ingress/egress through the CR doors. The normal HVAC system enables the release to the auxiliary building from containment. The exhaust through the plant vent stack for this model element is 12,290 cfm, as 10 cfm goes into the CR through the doors. (revised in total from the LAR submittal)

FUEL HANDLING ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-001, Version 3 Method/Computer Program Used: RADTRAD Version 3.03 Regulatory Guidance: RG-1.183, including Appendix B E1-3 to NL-17-1067 SNC Response to Request for Additional Information Model Discussion The calculation was performed to address a fuel handling accident (FHA) in the containment and in the SFP area of the Auxiliary Building.

For the containment accident, three combinations of containment equipment hatch and personnel airlock configurations are considered:

Open containment equipment hatch and open personnel airlock;

Closed containment equipment hatch and open personnel airlock; and

Open containment equipment hatch and closed personnel airlock.

If the containment equipment hatch or the personnel airlock is presumed to be open, no credit is taken to close it. The off-site doses calculated for the open containment equipment hatch and open personnel airlock bound the other two possible configurations. The open personnel airlock could allow areas around the Control Room (CR) to become contaminated, so the calculation accounts for dose impacts of ingress/egress through the CR doors. Also, a small amount of CR envelope wall internal to the Auxiliary Building is only 1 foot thick, so the shine from the contaminated area through the wall is added to the CR operator dose. Doses in the CR are accumulated over a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Releases from the damaged fuel are completed in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

For the accident in the SFP area of the Auxiliary Building, the accident releases also are completed in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The activity is released to the environment through the plant vent stack, and credit is taken for filtration of the iodine isotopes through the Penetration Room Filtration System. Doses from this accident are bounded by the doses from an accident in containment.

Results and Acceptance Limits EAB LPZ Control Room Release (rem TEDE) (rem TEDE) (rem TEDE)

Containment, EH &

2.4 0.9 1.7 PAL Open Bounded by Containment, EH Bounded by EH &

EH & PAL 2.3 Closed & PAL Open PAL Open Open Bounded by Containment, EH Bounded by EH &

EH & PAL 1.1 Open & PAL Closed PAL Open OQ_en Spent Fuel Pool 0.5 0.2 0.1 Acceptance Limit 6.3 6.3 5 (Note that round1ng 1s apphed to all values)

E1-4 to NL-17-1067 SNC Response to Request for Additional Information Key Assumptions and Inputs Source Term Parameters Parameter Value Reactor Power Level 2775 MWt (+2% uncertainty= 2831 MWt)

Reactor Peaking Factor 1.7 Fuel Movement Time 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months post shutdown.

Number of Fuel Assemblies 157 Number of Damaged Assemblies 1 Number of Damaged Fuel Rods 264 Table 1 - Core Cycle-to-Cycle Augments Isotope Factor Kr-85 1.15 Xe-133 1.05 Other Noble Gases 1.03 Other Iodine 1.03 isotopes Table 2 - Core Source Term Isotope Core Activity at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months post Shutdown (curies}

Kr-85 7.2E+05 Xe-131m 8.1E+05 Xe-133 1.0E+08 Xe-133m 2.0E+06 Xe-135 2.0E+05 1-131 5.4E+07 1-132 4.6E+07 1-133 5.7E+06 1-135 4.1E+03 E1-5 to NL-17-1067 SNC Response to Request for Additional Information Fraction of Fission Product Inventory in Gap 1-131 0.08 Kr -85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals (retained in pool) 0.12 Iodine Chemical Forms released from Fuel to Pool Elemental 99.85%

Organic 0.15%

Overlaying Pool Depth 23 feet Decontamination Factors Noble Gases 1 Particulates Infinity Overall Iodine 200 Iodine Chemical Forms Released from Pool Elemental 57%

Organic 43%

The released activity is obtained by making the product of the100-hour post-shutdown core activity for the isotope, the design margin, the gap fraction, and the radial peaking factor. The product is then divided by the Decontamination Factor (DF) and by 157 (the number of assemblies) to achieve the released activity shown in the last column of the table, below.

Table 3 - Net Scrubbed Release Activities 100-hr Radial Peaking Released Core Design Gap Group Isotope Factor OF Activity Inventory Margin Fraction (Curies)

(Ci)

Kr-85 7.20E+05 1.15 0.1 1.70E+00 1 8.97E+02 z Xe-0 8.10E+05 1.03 0.05 1.70E+00 1 C" 131m 4.52E+02 (i'

C) Xe-133 1.00E+08 1.05 0.05 1.70E+00 1 5.68E+04 m Xe-Ul (I) 2.00E+06 1.03 0.05 1.70E+00 1 Ul 133m 1.12E+03 Xe-135 2.00E+05 1.03 0.05 1.70E+00 1 1.12E+02

t: 1-131 5.40E+07 1.03 0.08 1.70E+00 200 2.41 E+02 m

0 1-132 4.60E+07 1.03 0.05 1.70E+00 200 1.28E+02 CQ (I) 1-133 5.70E+06 1.03 0.05 1.70E+00 200 1.59E+01

s Ul 1-135 4.10E+03 1.03 0.05 1.70E+00 200 1.14E-02 E1-6

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information Release from FHA in Containment Parameter Value Containment Volume 2.03E6 cubic feet Mixing Volume in Containment 1.0E6 cubic feet Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Containment Hatch Flow Rate 55,000 cfm Containment Release Filtration 0%

Auxiliary Building Mixing Volume 100,650 cubic feet Personnel Airlock Release Filtration 0%

Open Equipment Hatch & Open Personnel Airlock Case Off Site Doses (EAB and LPZ) Model Containment Hatch Flow Rate 55,000 cfm Personnel Airlock Flow Rate (see note below) 25,000 cfm Control Room Ingress/Egress Dose Model Containment Hatch Flow Rate 43,500 cfm Auxiliary Building Mixing Volume 100,650 cubic feet Personnel Airlock Flow Rate (see note below) 12,300 cfm Control Room Containment Hatch Dose Model Containment Hatch Flow Rate 55,000 cfm Personnel Airlock Flow Rate (see note below) 12,300 cfm Control Room Personnel Airlock Dose Model Containment Hatch Flow Rate 43,500 cfm Personnel Airlock Flow Rate (see note below) 25,000 cfm Open Equipment Hatch & Closed Personnel Airlock Case Containment Hatch Flow Rate 55,000 cfm Closed Equipment Hatch & Open Personnel Airlock Case Control Room Ingress/Egress Dose Model Auxiliary Building Mixing Volume 100,650 cubic feet Personnel Airlock Flow Rate (see note below) 12,300 cfm Control Room Personnel Airlock Dose Model Personnel Airlock Flow Rate (see note below) 25,000 cfm Note about Personnel Airlock flow rates: For the three model elements where release to the environment should be conservatively maximized, the flow rate through the Personnel Airlock is assumed to be 25,000 cfm and passes directly to the environment through the Plant Vent Stack as drawn by the normal auxiliary building HVAC system. This flow rate assumption conservatively increases the release to the environment and back into the CR through the CREFS intakes. For the model elements where the ingress/egress through the CR doors is modelled, the flow out the airlock into the auxiliary building hall is 12,300 cfm, which is based on sensitivity studies and the capacity of the normal HVAC system. The studies show that this flow rate maximizes the auxiliary building radioactivity concentrations, and causes the highest dose from ingress/egress through the CR doors. The normal HVAC system enables release to the auxiliary building from containment. The exhaust through the plant vent stack for this model element is 12,290 cfm, as 10 cfm goes into the CR through the doors.

E1-7

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information Release from FHA in the Spent Fuel Pool Area Parameter Value Fuel Handling Volume 72,150 cubic feet Overlaying Pool Depth 23 feet Fuel Handling Area Release Rate 5,000 cfm PRF Filtration 89.5% for iodine isotopes CR Parameters Parameter Value CRVolume 114,000 ft 3 CR Isolation Mode Initiation Automatic at 60 Seconds CR Pressurization Mode Initiation Manually at 21 minutes (20 minutes after isolation)

CR Ventilation System Normal Flow Rate 2340 cfm < 60 seconds CR Ventilation Isolation Mode Flow Rate 600 cfm (1 minute to 21 minutes)

CR Ventilation Pressurization Makeup Rate 375 cfm > 21 minutes CR Ventilation System Recirculation Flow Rate 2700 cfm > 21 minutes CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% all iodine species CR Pressurization Mode Unfiltered In-leakage 325 cfm*

CR Ingress/Egress Unfiltered In-leakage 10 cfm throughout (location changes)

CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 CR Ventilation Summarv Table 4 -Control Room Ventilation Summary Time Filtered Flow Unfiltered Flow (cfm) (cfm) 0 to 1 minute 0 2340 1 minute to 21 minutes 0 600 21 minutes to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 375 325

For the FHA in Containment, the 10 cfm for ingress and egress to the CR goes from the Auxiliary Building through the CR door. For the FHA in the spent fuel area, the 10 cfm for ingress and egress is conservatively added to the CR through the ventilation system and is unfiltered. This unfiltered in-leakage starts at time 0 and continues through the entire accident (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

E1-8 to NL-17-1067 SNC Response to Request for Additional Information Atmospheric Dispersion Factors (sec/m 3 )

Containment Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 8.79E-04 2-8 1.10E-4 6.77E-04 8-24 3.32E-04 Plant Vent Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 1.62E-03 2-8 1.10E-4 1.37E-03 8-24 7.10E-04 E1-9

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information RAI No. 37 (TSTF-448)

Regulatory Basis numbered 2, 4, 5, 6, and 13 apply to RAI No. 37.

TSTF-448, Revision 3, "Control Room Habitability," modifies Condition B and Condition F of Standard Technical Specifications (STS) Limiting Conditions of Operation (LCO) 3.7.10, "Control Room Emergency Filtration System (CREFS),"

in NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," to state:

Condition B One or more CREFS trains inoperable due to inoperable CAE boundary in MODE 1, 2, 3, or 4.

Condition F Two CREFS trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SNC has chosen to deviate from TSTF-448 by proposing to revise Condition B in FNP Technical Specification 3.7.1 0 so that it includes all modes of applicability instead of just Modes 1, 2, 3, and 4 and by not revising Condition D. Proposed Condition B will state:

One or more CREFS trains inoperable due to inoperable CAE boundary.

Current Condition D states:

Two CREFS trains inoperable in MODE 1, 2, 3, OR 4.

These two deviations from TSTF-448 cause an apparent conflict in FNP Technical Specification 3.7.10, such that two conditions are required to be entered anytime the following exist:

Two CREFS trains are inoperable because of CAE boundary while in Modes 1, 2, 3, or 4. This inoperability requires entry into Conditions B and D.

Two CREFS trains are inoperable because of CAE boundary during core alterations. This inoperability requires entry into Conditions B and F.

Two CREFS trains are inoperable because of CAE boundary during movement of irradiated fuel assemblies. This inoperability requires entry into Conditions B and F.

In addition, in letter dated June 18, 2009 (ADAMS Accession No. ML091690643),

the Technical Specifications Task Force (TSTF) submitted Traveler TSTF-508, Revision 1, "Revise Control Room Habitability Actions to Address Lessons Learned from TSTF-448 Implementation," to the NRC staff for review and approval. TSTF-508 proposed the extension of the use of mitigating actions to Modes 5, 6, and during movement of recently irradiated fuel assemblies when one or more CREFS trains is inoperable due to an inoperable CAE boundary in Westinghouse STS 3.7.10, just as SNC has requested through the deviation in E1-10

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information this LAR. During the review of TSTF-508, the NRC staff requested additional information (ADAMS Accession No. ML110890817). This position is applicable to this change requested by SNC and is as follows. The NRC staff expressed their view that the extension of the use of mitigating actions to Modes 5, 6, and during movement of recently irradiated fuel assemblies is not adequately justified and is not warranted for the following reasons:

The regulation at Subpart H of 10 CFR Part 20, "Standards for Protection against Radiation," provides the requirements for respiratory protections and controls to restrict internal exposure in restricted areas. Specifically, 10 CFR 20.1701 states that licenses shall use, to the extent practicable, process or engineering controls to control the concentration of radioactivity in the air. Use of other controls as described in 10 CFR 20.1702 is only allowed by regulation when it is not practicable to apply process or other engineering controls.

NEI 99-03, Appendix F, "Compensatory Measures Allowable On An Interim Basis," Page F-1, states:

The use of SCBA [self-contained breathing apparatus] and Kl [potassium iodide] has been determined to be acceptable for addressing control room envelope integrity in the interim situation until the licensee remediates the issue. However, use of SCBA or Kl in the mitigation of situations where in-leakage does not meet design basis limits is not acceptable as a permanent solution.

10 CFR 20.1701 essentially says that engineering/process controls shall be used to the extent practical. If not practical. then 10 CFR 20.1702 methods should be used.

Therefore. the use of SCBAs should be a last resort.

[emphasis added]

The use of Kl and SCBA is not without risk. The allowance to use Kl and SCBA was not previously extended to Modes 5 and 6 because another practical control (stopping fuel movement) existed. The NRC staff does not believe that the proposed compensatory measures are appropriate given that the process control of stopping fuel movement is available.

Explain how the regulations at 10 CFR 20.1701 and 10 CFR 20.1702 and the regulatory guidance in NUREG-1431 STS 3.7.10 are met or how your proposal meets the intent of these regulations and regulatory guidance. Please justify the deviation from TSTF-448 or remove the extension of the use of mitigating actions to Modes 5, 6, during core alterations, and during movement of irradiated fuel assemblies.

E1-11

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information SNC Response to RAI No. 37:

The standard TS used for TSTF-448 links the operability of CREFS and the Control Room Envelope (CRE). However, FNP's current TS 3.7.1 0 does not directly link CREFS and CRE operability.

FNP's TS 3.7.1 0 was last amended on September 30, 2004 (ML042780424).

That amendment re-titled TS 3.7.1 0 from CREFS to "Control Room," and the Safety Evaluation stated: The integrity of the CRE does not affect the operability of the CREFS ... An inoperable CRE most likely will have no effect on the capability of the CREFS ... "1 For this reason, FNP's existing TS does not, unlike the standard TS assumed by TSTF-448, have the limitations of "in Modes 1, 2, 3, or 4" for Condition B and for reasons other than Condition B" for Condition D.

To resolve the conflict identified in RAI No. 37, above, substituted TS mark-up and clean-typed pages are enclosed. The substituted TS pages will add "in Modes 1, 2, 3, and 4" to Action B and for reasons other than Condition B" to Action D. They also modify Condition F to properly reflect TSTF-448 changes.

The substituted pages2 will result in FNP's TS 3.7.10 being consistent with TSTF-448. The changes are not a deviation from NUREG-1431 and TSTF-448 because with these changes, the Farley TS will link CREFs and CRE operability.

For these reasons, 10 CFR 20.1701,1 0 CFR 20.1702 and the regulatory guidance in NUREG-1431 STS 3.7.10 are met.

Note that simultaneous entry into two Conditions will still be necessary (A and D),

if two CREFs trains are inoperable for reasons other than inoperable CRE boundary. However, this is entirely proper per the rules of the STS.

Finally, there is no need to amend the proposed changes to FNP's TS Bases sent with SNC's original submittal. Those original, proposed changes to the Bases are consistent with the substituted TS markup, since the original proposed changes to the Bases removed the language stating that CRE and CREF operability are independent of each other.

RAI No. 40 (RG 1.183)

Regulatory Basis numbered 1, 2, 3, 7, 8, and 9 apply to RAI No. 40.

Section 10 CFR 50.67(b}(2) requires that the licensee's analysis demonstrates with reasonable assurance that adequate radiation protection is provided to 1 See "Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments,"

ML042780424, Enclosure 3, "Safety Evaluation" at section 3.2.2, pages 9-10.

2 The markups and clean pages forTS 3.7.10-1, 3.7.10-2, and 3.7.10-3 are being substituted. The markup and clean page forTS 3. 7.10-4 remains the same as originally submitted. Please note that Surveillance Requirement SR 10.7.3 which appears at the bottom of the markup of page TS 3.7.10-3 appears on the clean page of TS 3. 7.1 0-4.

E1-12

Enclosure 1 to NL-17-1067 SNC Response to Request for Additional Information permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

In the LAR, SNC has provided the resultant radiation dose associated with occupancy of the control room. However, the LAR appears to be missing a discussion and/or calculation that accounts for the control room personnel radiation exposure received, for the duration of the accident, upon ingress/egress from the site boundary to the control room. In order to meet 10 CFR 50.67 and 10 CFR 50 Appendix A GDC 19, the radiation dose for accessing the control room must be evaluated from the site boundary to the control room for both ingress and egress for the duration of the accident.

Please provide an analysis of the radiation dose received from ingress and egress to the control room in enough detail that will enable the NRC staff to be able to perform an independent calculation.

SNC Response to RAI No. 40:

SNC has conservatively evaluated the dose that an operator would receive during a LOCA from ingress and egress to the control room during the entire 30 day event. The assumption inherent in the transit dose is that the operator travels to the site, parks in the parking lot, and walks from the parking lot to the Auxiliary Building entrance without any protection. The operator leaves using the same path. The transit dose calculation used the previously defined LOCA dose evaluation models and numerous other inputs and assumptions, detailed below.

The scenario is not the most likely scenario, as it ignores the principal of ALARA, but it represents a worst-case situation.

The dose to the operator was evaluated to be 0.2 REM. This is less than 5% of the total anticipated operator dose in the LOCA scenario. Contributions to the dose occur from containment leakage, ESF system leakage, RWST back-leakage, ground shine from deposited radioactivity, and containment shine.

Design Inputs/Assumptions:

1. Meteorological data is the same as used for the LOCA dose consequences analysis previously discussed in the submittal.

2. Containment leakage, quantity and timing, is the same as modelled in the LOCA dose consequences analysis.

3. ESF Leakage outside containment, quantity and timing, is the same as modelled in the LOCA dose consequences analysis (40,000 cc/hr is modelled in the operator transit dose analysis).

4. RWST leakage, quantity and timing, is the same as calculated in the LOCA dose consequences calculation. This iodine isotope leakage comes from the 2 gpm back leakage from RHR to the RWST modelled in the analysis.

5. Containment elevation is 290'3" (ground elevation is 154'6").

Containment is 65' in diameter, and the wall is 3'9" thick concrete with a

~' steel plate liner.

E1-13 to NL-17-1067 SNC Response to Request for Additional Information

6. Walking distance from the parking lot to the entrance to the control building is conservatively assumed to be 1501.5 feet.

7. The operators will be confined to the control room for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (100% occupancy factor) following the accident. The first ingress/egress will begin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter and be repeated every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for the length of the accident (30 days) (59 trips total).

8. Operator walking speed is 3 miles per hour (mph).

9. Operator respiration rate is twice the normal respiration rate for the control room (7.0E-04 m3/sec).

10. For the ground shine dose, 100% of the non-noble gas released activity is deposited on the ground of the plant site.

11. The dose point for the operator walking along the ingress/egress path is 5 feet off the ground.

12. Shielding provided by intervening equipment, plant buildings, or other materials is ignored (except for the containment concrete and liner when determining what the containment shine does to the operator).

13. Five points along the ingress/egress path are used to determine dose rates. (See table 1 and Figure 1 for the location of these points)

a. Atmospheric dispersion factors (X/Qs) are calculated from existing met-data for these receptor points. The X/Qs are substituted into the RADTRAD model for the LOCA, instead of the LPZ X/Qs to calculate the dose from the radioactive plume.

b. The ground shine and containment shine calculations use the source term from the environment and containment compartments as calculated in the RADTRAD model for the LOCA.

Table 1: Transit segments and points.

Transit Starting End Location Segment Segment Location Distance (ft}

1 Parking Lot Dose Point 1 326.5 2 Dose Point 1 Dose Point 2 106.1 3 Dose Point 2 Dose Point 3 220.6 4 Dose Point 3 Dose Point 4 436.4 5 Dose Point 4 Control Room 411.6 E1-14 to NL-17-1067 SNC Response to Request for Additional Information Figure 1 -Dose receptor point locations:

--* ~

I' Note, the radioactivity release points are indicated above, items 6, 7, and 8. All distances are straight line and no consideration is given to alternative pathways due to obstructions from other buildings. Although it is shown in the center of containment, the release from point 7 is actually taken from the surface of containment closest to each receptor point.

E1-15 to NL-17-1067 SNC Response to Request for Additional Information Table 2: X/Q dispersion factors:

95tb percentile ]JQ (s/m3 )

Receptor Release Location 4 to30 Location 0-2 hours 2-8 hours 8-24 hours 1 to 4 days days Unit 1 RWST 6.81E-05 6.23E-05 3.04E-05 2.68E-05 2.11E-05 Dose 1 Unit 1 Contaimuent 8.27E-05 7.73E-05 3.64E-05 3.23E-05 2.70E-05 Unit 1 Vent 9.19E-05 7.61E-05 3.65E-05 3.13E-05 2.55E-05 Unit 1 RWST 7.52E-05 7.01E-05 3.39E-05 2.93E-05 2.30E-05 Dose2 Unit 1 Contaimuent 9.74E-05 8.78E-05 4.32E-05 3.81E-05 3.13E-05 Unit 1 Vent 1.06E-04 8.90E-05 4.33E-05 3.69E-05 2.98E-05 Unit 1 RWST 9.72E-05 8.89E-05 4.28E-05 3.55E-05 2.73E-05 Dose 3 Unit 1 Contaimuent 1.41E-04 l.27E-04 6.34E-05 5.48E-05 4.32E-05 Unit 1 Vent l.51E-04 l.28E-04 6.36E-05 5.42E-05 4.30E-05 Unit 1 RWST 3.15E-04 2.79E-04 1.38E-04 l.05E-04 7.84E-05 Dose4 Unit I Contaimuent 7.95E-04 7.03E-04 3.36E-04 2.96E-04 2.49E-04 Unit 1 Vent 5.90E-04 5.00E-04 2.45E-04 2.02E-04 1.57E-04 Unit 1 RWST 4.17E-04 3.37E-04 1.50E-04 1.13E-04 8.71E-05 Dose 5 Unit I Contaimuent 2.28E-03 1.62E-03 7.32E-04 6.22E-04 4.64E-04 Unit 1 Vent I.20E-03 1.03E-03 5.12E-04 4.06E-04 2.67E-04 Table 3: Release Coordinates:

Release Point b. East (ft) b. North (ft)

Unit 1 RWST 232 -73.5 Unit 1 Containment 0 -140.5 Unit 1 Vent 44.8 -81 Table 4: Dose Pont Location Dose Location b. East(ft) b. North (ft) 1 -245 -1090 2 -260 -985 3 -300 -771.75 4 -71.5 -400 5 -150 -140.5 E1-16

Joseph M. Farley Nuclear Plant Units 1 and 2 Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request Enclosure 2 FNP Unit 1 and Unit 2 Technical Specifications Clean-Typed Pages

CREFS 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Filtration/Pressurization System (CREFS)

LCO 3.7.10 Two CREFS trains shall be OPERABLE.

NOTE ---------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train A.1 Restore CREFS train to 7 days inoperable for reasons OPERABLE status.

other than Condition B.

B. One or more CREFS B.1 Initiate action to implement Immediately trains inoperable due to mitigating actions.

inoperable CRE boundary in MODE 1, 2, AND 3, or4.

B.2. Verify mitigating actions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to 90 days OPERABLE status.

Farley Units 1 and 2 3.7.10-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

CREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1 21 31 I

or4.

C.2 ------------NOTE----------

LCO 3.0.4.a is not applicable when entering MODE4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Two CREFS trains D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable in MODE 1 2 1 I

31 OR 4 for reasons other AND than Condition B.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A not recirculation mode.

met during movement of irradiated fuel assemblies OR or during CORE ALTERATIONS. E.2.1 Suspend CORE Immediately ALTERATIONS.

AND E.2.2 Suspend movement of Immediately irradiated fuel assemblies.

Farley Units 1 and 2 3.7.1 0-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

CREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREFS trains F.1 Suspend CORE Immediately inoperable during ALTERATIONS.

movement of irradiated fuel assemblies or during AND CORE ALTERATIONS .

F.2 Suspend movement of Immediately irradiated fuel assemblies.

One or more CREFS trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies or during CORE ALTERATIONS .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the In accordance with heaters operating and each CREFS Recirculation and the Surveillance Filtration train for ~ 15 minutes. Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP Farley Units 1 and 2 3.7.10-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Joseph M. Farley Nuclear Plant Units 1 and 2 Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request Enclosure 3 FNP Unit 1 and Unit 2 Technical Specifications Marked-Up Pages

Control RoomCREFS 3.7.10 3.7 PLANT SYSTEMS

3. 7.10 Control Room Emergency Filtration/Pressurization System (CREFS)

LCO 3.7.10 Two Control Room Emergency Filtration/Pressurization System (CREFSj trains and the Control Room Envelope (CRE) shall be OPERABLE.

N0 T E ----------------------------------------------

The control room envelope (CRE ) boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS !

CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train A.1 Restore CREFS train to 7 days inoperable for reasons OPERABLE status.

other than Condition B.

B. One or more CREFS B.1 Initiate action to Immediately trains inoQerable due to implement mitigating inoperable CRE actions.

inoperable boundary in MODE 1. 2. 3. or4. AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.2. ~ Restore CRE to OPERABLE status.

GR B.2 .2. ~ General blesign Criteria 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

~ (GQC) ~ Q met using Verify mitigating actions mitigating actions in 8 .1.

ensure CRE occupant exposures to radiological, ANG chemical, and smoke AND hazards will not exceed limits. B . ~~ Restore CRE boundary to JG90 days OPERABLE status.

Farley Units 1 and 2 3.7.10-1 Amendment No. +99 (Unit 1)

Amendment No. +58 (Unit 2)

Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4.

C.2 ----------NOTE---------

LCO 3.0.4.a is not applicable when entering MODE4.

I Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Two CREFS trains D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable in MODE 1, 2, 3, OR 4 for reasons other AND than Condition B.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A not recirculation mode.

met during movement of OR irradiated fuel assemblies or during CORE ALTERATIONS. E.2.1 Suspend CORE Immediately ALTERATIONS .

AND E.2.2 Suspend movement of Immediately irradiated fuel assemblies.

Farley Units 1 and 2 3.7.10-2 Amendment No. +00 (Unit 1)

Amendment No . .:t-a8 (Unit 2)

Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. ReEjl::lireEl ,1\stion anEl F.1 Suspend CORE Immediately assosiateEl Com13letion ALTERATIONS.

+ime of ConElition not e met El1::1rin~ FRO¥ement of AND irraEliateEl f~::~el assemblies or El1::1rin§ CORe F.2 Suspend movement of Immediately Al+ERATIONS . irradiated fuel assemblies.

OR Two CREFS trains I OR I

I inoperable during movement of ir~

assemblies or durin

~ One or more CREFS trains inoperable due to an CORE AL~~~ ";-110NS.

inoperable CRE boundary during movement of irradiated fuel assemblies or during CORE ALTERATIONS .

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the In accordance with heaters operating and each CREFS Recirculation and the Surveillance Filtration train for ~ 15 minutes. Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). VFTP SR 3.7.10.3 ------------------------NOTE------------------------------ In accordance with Not required to be performed in MODES 5 and 6. the Surveillance Frequency Control Verify each CREFS train actuates on an actual or Program simulated actuation signal.

Farley Units 1 and 2 3.7.10-3 Amendment No. ~ (Unit 1)

Amendment No. 4-W (Unit 2)

v t e Letter Sequence Response to RAI
CAC:MF8861, (Approved, Closed)
Initiation Request, Request, Request Acceptance Supplement, Supplement Administration Questions 24 March 2017 21 July 2017 Answers 23 May 2017 8 June 2017 Results Approval Other:
MONTHYEAR2017-03-24 [Table View]

NL-17-1067, Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request

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NL-17-1067, Response to Request for Additional Information Regarding Alternative Source Term License Amendment Request

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