Text

Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 6
	ML16071A147
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	03/03/2016
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16071A108	List: ML16071A110; ML16071A126; ML16071A132; ML16071A136; ML16071A137; ML16071A140; ML16071A144; ML16071A146; ML16071A147; ML16071A149; ML16071A150; ML16071A152; ML16071A153; ML16071A158; ML16071A160; ML16071A166; ML16071A170; ML16071A174; ML16071A176; ML16071A180; ML16071A181; ML16071A184; ML16071A185; ML16071A187; ML16071A188; ML16071A193; ML16071A195; ML16071A198; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 1 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 2 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 3 of 4; NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 4 of 4; NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 3; NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 3; NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 3; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 6; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 6; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 6; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 6; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 6; NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 6; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 7 of 8; NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 8 of 8; NL-15-1898, Enclosure 5: Vogtle Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 4; ... further results
References
	NL-15-1898
	Download: ML16071A147 (16)
	v • d • e

Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant Units 1 and 2;Edwin I. Hatch Nuclear Plant Units 1 and 2;Vogtle Electric Generating Plant Units 1 and 2;License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 4 EAL Verification and Validation Documents Joseph M. Farley Nuclear Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 4 Farley EAL Verification and Validation Documents Farley EAL V&V Documentation Index V&V# ~Descripjtion -IC i Threshold; RCS Modes/TemPeratures

-TS Reference (Table 1.1-1) EAL Basis Section 1.4 1 RCS Cold Shutdown Temperature reference (200 °F) CA3 (1)CU3 (1)Radiation Monitor Calculations:

RG 1 (1)RE-15C RS1 (1)RE-29B (NG) RA1 (1)Calc SM-SNC524602-O01 Radiation Monitor Calculations:

RUl1 (1)RE-18 RE-23B RE-15 R- 14 RE-22 RE-29B 2 Caic SM-SNC52 460 2-OO1 Radiation Monitor Calculations:

CG 1 (1)b RE-27A&B CS1 (2)b Caic SM-SNC524602-O01 Containment Radiation Monitor Calculations:

FCB L3A FCB L3A RE-27A&B CB PL3A CB PL3A Caic SM-SNC52 4602-O01 Containment Radiation Monitor Calculations:

RCB L3A RCB L3A RE-2 RE-7 Calc SM-SNC52 4602-O01 ODCM/TS Reference to Site Boundary RG1 (2)(3)3RS1 (.2)(3)RA1 (2)(3)(4)SFP Level 3 & 2 Indications RG2 (1)4 RS2 (1)RA2 (3)Alarm Response -FNP- 1(2)-ARP-1.6 RA2 (2)5 Annunciator Display: RE-24A&B RE-5 RE-25A&B____

Alarm Response -FNP- 1-ARP- 1.6 Control Room Rad RA3 (1)6 Monitor (Annunciator Display RE-l1A)______

Alarm Response -FNP-I(2)-ARP-I

.5 (SFP Level) RU2 (1)a Annunciator Display: EH2 (SFP LVL HI/LO)Radiation Monitor Information (FSAR): RE-5 RU2 (1)b 8 RE-2 RE-27A&B 9 RVLIS/RPV Level CS 1 (l)b FNP-1(2)-UOP-4.3 CAl (1)10 RPV Level Calculation CS1 (1)b Containment Sump/Reactor Coolant Drain Tank CG 1 (1)b 11 (RCDT)/Waste Holdup Tank (WHT) FSAR Reference CS 1 (2)b CAl (2)b CU1 (2)b H 2 Concentration

(> 6%) -calculation/reference CG1 (1)c 12 Calc SM-95-0754-001 CB PL4B CB PL4B 2/22/2016 Page 1 of 2 Farley EAL V&V Documentation Index V&V# Description IC Th reshold ESF Busses Drawing CA2 (I)CU2 (1)a SG1 (1)a 13 SG2 (1)a SS1 (1)SA1 (1)SU1 (1)DC Voltage Reference CU4 (1)14 SG8 (1)b SS8 (1)15 ISFSI TS/Dose Reading Calculation E-HU1 (1)Caic SM-SNC5 2460 2-O01 CSFST Information FCB PL1A FCB PL1A RCB PL1B RCB PL1B FCB L2A FCB L2A FCB PL2A FCB PL2A FCB PL2B FCB PL2B 16 RCB PL2A RCB PL2A CB PL2A CB PL2A CB PL4A CB PL4A CB PL4C. 1 CB PL4C. 1 SG1 (1)b 9S5 (1)c Rad Monitor Reference -FNP-I (2)-ARP-1.6 CB L4B CB L4B RE- 10 17 RE-14 RE-21 RE-22 Seismic ARP -FNP-l(2)-ARP-1.12 HU2 (1)18 (Annunciator Display (K5))______

RCS sample activity -Tech Spec Values SU3 (1)19 TS 3.4.16 RCS Leakage -Tech Spec Values SU4 (1)20 TS 3.4.13 ______ (2)21 Containment Spray Initiation Setpoint (27 psig) SU7 (2)a 2/22/2016 Page 2 of 2 Vl Page 1 of 1 Definitions 1.1 Table 1.1-1 (page 1 of 1)MODES MODE TITLE REACTIVITY

% RATED AVERAGE CONDITION THERMAL REACTOR COOLANT (keff) POWER(a) TEMPERATURE (0 F)1 Power Operation

_> 0.99 > 5 NA 2 Startup >_0.99 _<5 NA 3 Hot Standby < 0.99 NA _> 350 4 Hot Shutdown(b)

< 0.99 NA 350 > Tavg > 200 5 Cold Shutdown(b)

< 0.99 NA _<200 6 Refueling(C)

NA NA NA (a)(b)(c)Excluding decay heat.All reactor vessel head closure bolts fully tensioned.

One or more reactor vessel head closure bolts less than fully tensioned.

Farley Units 1 and 2 1.1-7 Amendment No. 146 (Unit 1)Amendment No. 137 (Unit 2) 7 V2 See NL-1 5-1 898 Enclosure 3 SIUTrHEiRI A, Southern Nuclear Design Calculation-Calculation Number: SM-SNC524602-001 Plant: IFarley Nuclear Plant Un:171-2 ]1& Dsilr"M hicl Title:

Subject:

Emergency NEI 99-0 1 Rev 6 EAt Calculations IAction Level Setpoints Purpose I Objective:

Document Emergency Action Level Values to support conversion to NEI 99-01 Rev 6 SSystem or Equipment Tag Numbers: NlA Contents Topic Page Attachments

of (Computer Printouts, Technical Papers, Pages_________________Sketches, Correspondence),Purpose 1 A-"SNC EP Concurrence 1 Criteria 1 B -Reserved 0 Conclusions

.....3 C -References 23 Design,, Inputs 6 'D" Water Level Elevations Corresponding to2 Assumptions 19 Fuel Uncovery2 References 22 E -TEDE & Thyroid CDE Dose Calculations 43 Method of Solutions 25 F -Shielding Calculations 29 Body o'f Calculation 30 G -Fuel Clad Barrier Threshold Caliculations 44_____________H

-RA1 Calculation 18 I- E-HU1 Calculation 16 Total # of Pages including cover sheet & 21 Attachments:-

Nuclear Quality Level 0 ' Safety-Related IZI Safety Significant 0] Non- Safety --Significant Version Record Veruion Originator Reviewer' Approval 1 Approval 2 No. Desciu o fc m .w o PI.t~Wo Nag PtWd d Calculation of EAL Thresholds C=.I -I to support NEI 99-01 Rev 6 1 r'JAO" ,..(e '.I'-'U I, NMP-ES-039-F01NM-S3901 NMP-ES-039-001 V3 Page 1 of 1 FNP-ODCM Figure 10-1 Site Map for Effluent Controls 10-7 Version 24 01/10 10-7 Version 24 01/10 V4 Page 1 of 3 Enclosure 2 to 5-0699 Joseph M. Farley Nuclear Plant -Unit 1 Responses to NRC Requests for Information The information provided below is generally a culmination of the previously provided SNC responses to NRC letter, Interim Staff Evaluation and Request for Additional Information

-Farley Nuclear Plant Units 1 and2 Regarding Overall Integrated Plan (OIP) for Reliable Spent Fuel Pool Instrumentation (Order EA.-12-05 1), dated October 30, 2013. Some information has been updated and/or revised including FIA #11 Unit 1 power supplies and other status completions.

Requests for Information (RAI) numbers 4, 10, and 14 were not included in the NRC letter dated October 30, 2013 and are identified as "Not Applicable" below.NRC RA! 1: Please provide clarification on the elevation identified as Level 2, the elevation of the highest point of any fuel rack seated in the SFP, and the elevation identified as Level 3.SNC Response to RAJ 1: L~evel 3 is the nominal Top of Rack elevation at 1 29'-3.5".

Level 2 is 10 feet above at elevation 1 39'-3.5".NRC RA! 2: Please identify the final SFP level instrumentation measurement range appropriate to the resolution of the Levels identified in the response to RAI #1 above.SNC Response to RAI 2: The SFP level lnstrumentation measurement range Is 1 53'-1 0.0" to 1 30'-1 .5" (+/- 1 ft. of the top of the fuel rack, per NEI 12-02).NRC P.A!3: Please provide a clearly labeled sketch or marked-up plant drawing of the plan view of the SFP area, depicting the SFP inside dimensions, the planned locations/placement of the primary and back-up SFP level sensor, and the proposed routing of the cables that will extend from these sensors toward the location of the read-out/display device.SNC Response to RAI 3: A plan view sketch of the SFP area is provided as Figure 1. The sketch depicts the SFP inside dimensions, locations/placement of the primary and alternate level sensors, and the routing of cables that extend toward the location of the electronics.

Physical separation of the primary and alternate channels, to the extent practicable and comparable to the short side of the pool, is used to provide reasonable protection of the level indication function against missiles that may result from the damage to the structure over the SFP.F_2-1 V4, Page 2of 3 Southern Nuclear Operating!

Company Nuclear SOr, TnuIERNA Management DCP Discipline J001 Worksheet Page 1 of 7~@MPANV Form Plant: Farley [] Hatch LI Vogltle Iii Unit No. 1 Iii 2 [] Shared Iii Worksheet No.: SNC467145J001 Worksheet Version No.: 2.0 For the applicable Discipline for this Worksheet, this DCP is a I- Partial Response Z] Complete Response.For changes to DCP Version 2.0, refer to section "Response to FCRs and/or later DCP versions" During the events at the Fukushima Daiichi Nuclear Power Station, responders were without reliable instrumentation to determine water level in the Spent Fuel Pool (SFP). This caused concerns that the pool may have boiled dry, resulting in fuel damage. While it was eventually determined that the SFP integrity and fuel cooling were not compromised, considerable efforts were made at the time to assure adequate cooling of the fuel in the SFP. These efforts diverted resources and attention from other critical tasks required to manage the event. As a result, the Nuclear Regulatory Commission (NRC)issued order EA-12-051 requiring plants to provide a reliable means of determining SFP level so that responders would have reliable information of the conditions of the SEP.This DCP will install a non-safety Spent Fuel Pool Level Instrumentation System (SFPLIS) that uses guided wave radar technology.

Two separate channels (primary and alternate) of level measurement will be installed.

Each channel will include the following:

Component Primary Channel Location Alternate Channel Location Probes Unit 2 Spent Fuel Pool Room 2240 Unit 2 Spent Fuel Pool Room 2240 (G3LE07&El I155'-0" El 1 55'-0" (N2G31 LE00078&Transmitters Unit 2 MCC Room 2478, El 155'-0" Unit 2 MCC Room 2478, El 155'-0" (N2G31LT0007

&N2G 31 LT0008)Electronic Enclosures Unit 2 Clean Storage Area Room Unit I Hallway Room 345, El 1 39'-0" (N2G31 L0007 & 2452, El 1 55'-0" N2G31L0008) with display (N2G31LI0007

& 0008)A brief overview of these components is as follows: Probe -the probe is a 5 mm flexible 316 stainless steel cable with a counter weight attached to the bottom. A mount bracket positions the probe inside the pool and attaches the probe assembly to the pool deck. A coupling transitions and electrically connects the probe to a coaxial cable, which electrically connects the probe to the transmitter.

For connection and mounting details refer to worksheets J003, E001, C001 Transmitter

-the transmitter generates and times the pulse that is transmitted to the probe to measure pool level. The transmitter measures the return time of the pulse and calculates the level. A 4-20mA current loop signal cable communicates the level data to the electronic enclosure.

For connection and mounting details refer to worksheets J003, J01 7, J01 8, E001, CO0l Electronics Enclosure

-the electronics enclosure receives the 4-20mA current loop signal from the transmitter, processes the signal, and displays the pool level on its front panel. For connection and mounting details refer to worksheets J003, J023, E001, CO0l NMP-ES-O44-F07, Version 2.0 NMP-ES-044 V4 Page 3 of 3 Spent Fuel Pit Cooling & Purification System FNP-1-SOP-54.0 FARLEY [ Version 72.0 Unit 1 Page 44 of 92 FIGURE 3 Page 1 of 1 SFP Level Probe Depths COAXIALCAL COUPLER Launch piato EL. 154' -1 SFP CUR8B EL, 15$' -0 Probe Io-,omu EL. 12W -SS"SENSOR PROBE SPENT FUEL ASSEMBLY 8o1t0iu of P.oI EL. 114' --rlnzeo a[

V5 S Page 1 of 10 07/13/15 09:23:59 L.

LOCATION FHIi SETPOINT:

1. Variable, as per FNP-1-RCP-252 ORIGIN: Any of the below listed Area, Process or Gaseous and Particulate Monitors:

R01, R02, R03, R06, R07, ROB, R10, R11, R12, R13, R14, R15, RI7A R17B, RIB, R-19, R20A, R20B, R21, R22, R23A or R23B.PROBABLE CAUSE HI RMS I .High Radiation Level in the System, Area or at the Component monitored.

2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTOMATIC ACTIONS!.The following actions will occur if a High Radiation Alarm is actuated on the associated Radiation Monitor.A. R14: (Plant Vent Gas) closes Waste Gas Release Valve 1 -G WD-HV-01 4.B. R16: None -abandoned in place per MDC 1040589501.

C. RI7A or B: (Component Cooling Water) closes Q1P17RCV3028 CCW SRG TANK VENT.D. RIB8: (Liquid Waste Processing) closes Liquid Waste Release Valve 1 -L WP-RC V-0 18.E. RI9: (Steam Generator Blowdown) isolates Steam Generator Blowdown Sample Lines.F. R23A :( Steam Generator Blowdown Processing) closes 1-BD-FC V-I 152 S/G Biowdown Heat Exchanger Outlet FCV.0. R23B :( Steam Generator Blowdown Processing) closes 1-BD-RCV-023B SQ BLOWDOWN PM? DISC.[Page 1 of 15 Page ofiSVersion 72.1 07/13/15 09:23:59 V5 Page 2 of 10 FNP-1-ARP-I

.6 LOCATION .Fill RADIATION MONITOR REFERENCE TABLE 1~ I I Y RE LOCATIONi DETECTOR FUNCTION ACTIONS R-I A Control Room (Unit I Area G-M (__W) Perform Step Panel) ___ _______4.1 R-I B Technical Support Center Area G-M ( _.W) No input to.......... (U nit II Panel) R-IB ___......._

_ this al.armi R-2 Containment (155' elev) Area G-M (_W ) Starts IPC dose Perform Steps_____integration 4.2 R-3 Radiochemnistry Lab (AB Area G-M ( _W) Perform Step_ _ _ _ 139') _ _ _ _ _ _ 4.3 R-4 #3 Charging Pump (AB Area G-M (_W ) Perform Step I___ _ 00) _ _ _ _ _ _ _ _ _ _ _ _ _ 4.4 R-5* Spent Fuel Pool Room (AB Area G-M (_.) Perform Steps 155') -______4.5 R-6 Sampling Room (AB 139') Area G-M ( _W) Perform Step____ __ ___ __ ___ ____ ___ __ ___ ___ ___ 4.6 R-7 In-core NIS Area (CTMT Area G-M ( __W) Starts TPC dose Perform Steps____129', near Seal Table) ___________integration 4.7 R-8 Drumming Station CAB Area G-M ( W) Perform Step_ _ _ _ 155') __ _ __ _ _ _ _ _ _ 4.8 R-9 SG Sample Panel (Unit UI Area G-M ( W) No input to____Panel) (AB 139') _ _________this alarm R-1 0 Penetration Room Filtration APD Beta Scint Perform Step____Discharge (AB 155') ____**(GA-ES)

_______4.9 R-1I

Containment Atmosphere APD Beta Scint Perform Step (AB 121 ') ____**(GA-.ES)

_______4.10

Technical Specification related**eea Atomics Electronic Systems Page 5 of 15 Vrin7.Version 72.1 V5 Page 3 of 10 FNP-I -ARP-1 .6 07/1 3115 09:23:59 LOCATION .HIII RADIATION MONITOR REFERENCE TABLE (cont)RE LOCATION fTYE DETECTOR FUNCION~ ACTlION R-21 Plant Vent Stack (AB 155') APD Scint. Perform Step__________________(Victoreen)

_______4.20 R-22 Plant Vent Stack Gas G-M (W Perform Step ODCM (AB 155') ) 4.21 R-23A SG Blowdown Surge Tank Liquid Scint. ( W) Closes Perform Step____Inlet (AB 130') _ ____ FCV-1152 4.22 R-23B SO BLOWDOWN PMP Liquid Scint. ( W) Closes Perform Step ODCM DISC (AB 130') ______RCV-23B 4.22 R-24A* Containment Purge (AB Gas Scint. Closes No input to 155') containment this alarm purge supply &exhaust dampers 2866C & 2867C__________________________and 3198A & D ____R-24B* Containment Purge Gas Scint. Closes valves: No input to (AB 155') (Victoreen) 2866D &2867D, this alarm 3196, 3197,____ ________3198B

&C _____R-25A/ Spent Fuel Pool Ventilation Gas Scint. Trip fuel bldg No input to R-25B* (AB 184') (Victoreen) supply and this alarm exhaust fans, closes SFP HVAC supply and exhaust dampers; starts associated trains of penetration

_______________room filtration.

R-27A* Containment (High Range) Area Ion Chamber No input to___ ___________(Victoreen) this alarm Technical Specification related Page 7 of 15 Vrin7.Version 72.1 V5~Page 4 of 10 07/13/15 09:23:59 i !, F'NV-l-ARP-I.6 LOCATION F4 SETPOINT:

Variable, as per FNP-l-RCP-252 ORIGIN: Radiation Monitor Cabinet Channels R-24A or R-24B Containment Purge PROBABLE CAUSE H4 CP RE24 A OR B HI RAD 1. High Radiation Level in the Containment Purge Exhaust Line. , i2. The radiation monitors fail to a "High Radiation" condition on oss of instrument and/or control power that will re~sult in actuation of associated automatic functions.

AUTOMATIC ACTION I. Isolates Containment by closing Purge Supply and Exhaust Valves I -CP-HV-3 196, l-CP-HV-3 197, I-CP-HV-3198A, B, C, & D, l-CP-HV-2867C

& D and 1-CP-HV-2866C

& D.OPERATO CION~1. Determine which radiation monitor indicates high activity.

r-]2. verif~y that any required automatic actions have occurred and if required, secure any running containment purge or mini-purge fans per FNP-l-SOP-12.2, CONTAINMENT PURGE AND PRE-ACCESS FILTRATION SYSTEM. El 3. Notify HP personnel of alarm. I-]4. Implement NMP-EP-l110 EMERGENCY CLASSIFICATION DETERMINATION AND INITIAL ACTION. C" 5. Determine the validity of the high activity indication as follows: 5.1 Verifyr that the instrument is aligned for normal operation and is functioning properly.

[]5.2 Sample or survey the affected system or area as required.

C]6. Determine the source or cause of the high activity and correct or isolate as required, I-]7. DO NOT allow personnel to enter the affected area without the approval of the Health Physics Department.

El 8. IF high activity indication is due to instrument failure, refer to Technical Specifications, section 3.3.6. C'9. IF high activity indication of RCS leakage is present AND accompanied by either decreasing pressurizer level OR decreasing VCT level, THENj go to FNP-I-AOP-1.0, RCS LEAKAGE. 0-10. 2WHI activity levels have decreased below the alarm setpoint, THEN reset the HI alarm on the RAD monitor drawer by depressing the FAIL/RESET pushbutton.

C]

References:

A-177100, Sb.. 309; U-258400; D-181658; D-181671; D-177199; 1D-177204; FSAR, Section 11.4; D.-175010, Sh. 2.Page 1 of I Page oflVersion 72.1I V5 Page 5 of 10 FNP-I -ARP-I .6 07/13/15 09:23:59 SETPOINT: Variable, as per FNP-I-RCP-252 LOCATION F.EH5 H5 SFP AREA RE25 A OR B HIRAD ORIGIN: Radiation Monitor Cabinet Channels R-25A or R-25B, Spent Fuel Pool Vent 1. High Radiation Level in the discharged air from the Spent Fuel Pool Area [2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTMATIC CTIN NOTE: The unaffected train penetration room filtration system may also start, due t NT: low M' in the spent fuel pool.I-Trips the Fuel Handling Area Supply and Exhaust Fans, closes the Fuel Handling Area Supply and Exhaust Dampers AND starts the Penetration Room IA OR lB Filtration Units.OPERATOR ACTION 1. Determine which radiation monitor indicates high activity.2. IF the alarm is due to a spike as indcated by the drawer ALERT light illuminated, THE check that the activity level has decreased below the alarm setpoint.2.1 IF the activity level has below the alarm setpoint, THEN reset the ALERT alarm on the RAD monitor drawer by depressing the FAIL/RESET pushbutton.

3. IF R25A in HIGH alarm, THEN verify open SEP TO lA PRE SUPPLY DMPR, Qi V48HV3538A.

4. IF R25B in HIGH alarm, THEN verify open SEP TO 1 B PRF SUPPLY DMPR, Q1V48HV3538B.

5. Verify that the required automatic actions listed above have occurred.

IF any automatic actions have not occurred, THEN1 go to FNP-I-SOP-5 8.0.(The section for Fuel Handling Area Heating and Ventilation Operation for guidance)6. Announce receipt of the alarm and the affected area on the public address system.El El LI LI-Page I of 2 Page of 2Version 72.1 iif~~ F)Page 6 of 10 LOCATION .fL SETPOINT:

1. Variable, as per FNP-2-RCP-252 ORIGIN: Any of the below listed Area, Process or Gaseous and Particulate Monitors:

R01B, R02, R044~,[--

R06, R07, R08, R09, RI0, RI 1, Rl2, R13, RI4, RI5, RI7A, RI7B, RIB, RI9, R2OA, R20B, R21, R22, R23A or R23B PROBABLE CAUSE RMS HI-RAD 1 .Hih Radiation Level in the System, Area or at the Cornonent monitored.

2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTOMATIC ACTION I. The following actions will occur if a High Radiation Alarm is actuated on the associated Radiation Monitor.a) R14: (Plant Vent Gas) closes Waste Gas Release Valve 2-GWD-HV-0 14.b) R16: (Boron Recycle System) diverts 2-C VC-RCV-016 Recycle Evaporator discharge from Reactor Makeup Water System to the Recycle Evaporator Demineralizer.

c) RI7A or B: (Component Cooling Water) closes 2-CCW-RCV-3028 CCW SRG TANK VENT.d) R 18: (Liquid Waste Processing) closes Liquid Waste Release Valve 2-LWP-RCV-0 18.e) RI9: (Steam Generator Blowdown) isolates Steam Generator Blowdown Sample Lines.0 R23A: (Steam Generator Blowdown Processing) closes 2-BD-FC V-i1152 S/G Blowdown Heat Exchanger Outlet FCV.g) R23B: (Steam Generator Blowdown Processing) closes 2-BD-RCV-023B SGBD DISCH TO ENVIRONMENT.

Page I of 13 Vrin6.Version 61. I 07113/15 09:41:15 Uy+/-IT 2 OPERATOR ACTION (conti V5 Page 7 of 10 FNP.2-ARP-1

.6 LOCATION .H 4. In addition to the general actions perform steps indicated in the "ACTIONS" column of the following table. (Some of the radiation monitors included in the table do not input into this alarm but are included for reference).

RADIATION MONITOR REFERENCE TABLE C" RE LOCATION fl FUNCTION ACTIONS R- IA Control Room (Unit I Area G-M ( W) No input to___Panel)

___ _____this alarm R- 1 B Technical Support Center Area G-M (YW ) Perform Step (Unit II Panel) R- 1B __ ____ ______4.1 R-2 Containment (155' elev) Area G-M ( W) Starts IPC dose Perform Steps________________

_____integration 4.2 R-3 Radiochcmistry Lab (AB Area G-M ( W) No input to 139') ___ _____this alarm R-4 #3 Charging Pump (AB Area G-M ( W) Perform Step_____100')____

4.3 R-5* Spent Fuel Pool Room (AB Area G-M ( W) Perform Steps_ _ _ _ 155') ________ 4.4 R-6 Sampling Room (AB 139') Area G-M ( W ) Perform Step____ ___ ____ ___ ____ ___ ___4.5 R-7 In-core NISArea (CTMT Area G-M ( W) Starts IPC dose Perform Steps 129', near Seal Table) ____integration 4.6 R-8 Drumming Station (AB Area G-M ( W ) Perform Step 155') ____4.7 R-9 SG Sample Panel (Unit II Area G-M ('W ) Perform Step Panel) (AB 139') ____4.8 R-10 Penetration Room Filtration APD Scint. Perform Step Discharge (AB 155') ___ (Victoreen) 4.9 R-II " Containment Atmosphere APD Scint. Perform Step____(AB 121') ___ (Victoreen)

_______4.10

Technical Specification related Page 4 of 13 Page 4of 13Version 61.1 UNIT 2 V5 Page 8 of 10 FNP-2-ARP-1.6 07113115 09:41:15 LOCATION Fi.Ell RADIATION MONITOR, REFERENCE TABLE (cont)RE j fLOCA IO TYP DETECTOR ACTIONS~R-20B Service Water from Liquid Scint. ( W) Perform Step Containment Coolers C and 4.19 D (AB 121') ___ _____R-21 Plant Vent Stack (AR 155') APD Scint Perform Step_______(Victoreen)

_______4.20 R-22 Plant Vent Stack Gas G-M ( W) Perform Step ODCM (AB 155') ___________

______ 4.21 R-23A SG Blowdown Surge Tank Liquid Scint. ( W) Closes Perform Step Inlet (AD 130') ____ ______ FC V-1152 4.22 R-23B SGBD DISCH TO Liquid Scint. ('W ) Closes Perform Step ODCM ENVIRONMENT RCV-23B 4.22 (AB 130')_______

R-24A* Containment Purge (AB Gas Seint. Closes No input to 155') containment this alarm purge supply &exhaust dampers 2866C & 2867C_______ ___ ___ __ ___and 3198A &D R-24B* Containment Purge Gas Scint. Closes valves: No input to (AB 155') (Victoreen) 2866D &2867D, this alarm 3196, 3197,_________________

_______3198B

&C R-25A1 Spent Fuel Pool Ventilation Gas Scint. Trip fuel bldg No input to R-25B* (AB 184') (Victoreen) supply and this alarm exhaust fans;closes SFP HVAC supply and exhaust dampers; starts associated trains of penetration room filtration.

Technical Specification related Page 6 of 13 Vrin6.Version 61. !

V5 U N LPage 9 ofo 10 07/13/15 09:41:15 ' NFNP-2-ARP-1.6 LOCATION FH4 SETPOINT:

Variable, as per FNP-2-RCP-252 H4j CP ORIGIN: Radiation Monitor Cabinet Channels R-24A or RE24 A OR B R-24B Containment Purge HI RAD PROBABLE CAUSE 1 .High Radiation Level in the Containment Purge Exhaust Line.2. The radiation monitors fail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTOMATIC ACTION I. Isolates Containment by closing Purge Supply and Exhaust Valves 2-CP-HV-3 196, 2-CP-HV-3 197, 2-CP-H-V-31I98A, B, C, & D, 2-CP-HV-2867C

& D and 2-CP-HV-2866C

& D.OPERATOR ACTION 1. Determine which radiation monitor indicates high activity.

El 2. Verify that any required automatic actions have occurred and if required, secure any running containment purge or mini-purge fans per FNP-2-SOP-12.2, CONTAINMENT PURGE AND PRE-ACCESS FILTRATION SYSTEM. []3. Notify HP personnel of alarm. IL]4. Implement NMP-EP-1 10 EMERGENCY CLASSIF'ICATION DETERMINATION AND INITIAL ACTION. I]5. Determine the validity of the high activity indication as follows: 5.1 Verify that the instrument is aligned for normal operation and is functioning properly.

[]5.2 Sample or survey the affected system or area as required.

LI 6. Determine the source or cause of the high activity and correct or isolate as required.

El 7. DO NOT allow personnel to enter the affcted area without the approval of the Health Physics Foreman. 0" 8. IF high activity indication is due to instrument failure, THEN refer to Technical Specifications, section 3.3.6.9. IF high activity indication of RCS leakage is present AND accompanied by either;9.1I Decreasing pressurizer level El 9.2 Decreasing VCT level 0" 10 THEN go to FNP-2-AOP-I.0, RCS LEAKAGE. LI 11I. WHEN activity levels have decreased below the alarm setpoint, THEN reset HI alarm on the RAD monitor by depressing the FAIL/RESET pushbutton.

E"l

References:

A-207100, Sh. 309; U-213901, D-204658; D-204671; D-207199; D-207204;FSAR, Section 11.4; D-205010, Sh. 2 Page 1 of !Page of IVersion 61.1I V5 UNIT 2 Paelof0o, 07/13/15 09:41:15 .wFNP-2-ARP-1

.6 LOCATION .I'IL SETPOINT:

Variable, as per FNP-2-RCP-252 "5 SFP AREA ORIGIN: Radiation Monitor Cabinet Channels R-25A or jRE25 A OR B R-25B, Spent Fuel Pool Vent jHI RAD PROBABL E CAUSE 1. High Radiation Level in the discharged air from the Spent Fuel Pool Area Ventilation Fans.2. The radiation monitors f'ail to a "High Radiation" condition on loss of instrument and/or control power that will result in actuation of associated automatic functions.

AUTOMATIC ACTION NOTE: el The unaffected train penetration room filtration system may also start, due to low AlP in the 1. Trips the Fuel Handling Area Supply and Exhaust Fans, closes the Fuel Handling Area Supply and Exhaust Dampers AND starts the Penetration Room 2A OR 2B Filtration Units.OPERATOR ACTION 1. Determine which radiation monitor indicates high activity.

El 2. IF the alarm is due to a spike as indicated by the drawer ALERT light illuminated, THEN check that the activity level has decreased below the alarm setpoint.

!"l 2.1 IF the activity level has decreased below the alarm setpoint, THENj reset the ALERT alarm on the PAD monitor drawer by depressing the FAJL'RE.SET pushbutton

['l 3. IF R25A in HIGH alarm, TljN verify open SIP TO 2A PRF SUPPLY DMPR, Q2V48HV3538A.

El 4. IF R25B in HIGH alarm, T J verify open SIP TO 2B PRF SUPPLY DMPR, Q2V48HV3538B.

I"]5. Verify that the required automatic actions listed above have occurred.

IF any automatic actions have not occurred, go to FNP-2-SOP-58.0. (Th section for Fuel Handling Area Heating and Ventilation Operation for guidance)

['6. Announce receipt of the alarm and the affected area on the public address system. !"l Page I of 2 Page of2Version 61I.1I 07113115 09:23:59 V6 Page 1 of 1 FNP-1-ARP-I,.6 LOCATION F..Eill RADIATION MONITOR REFERENCE TABLE LOCATION rn DETECTOR FUNCTION ACTIONS R-1IA Control Room (Unit I Area (3-M (_W ) Perform Step____Panel)

___ _______4.1 R-1 B Technical Support Center Area G-M ( WL) No input to___(Unit II Panel) R-IB this alarm R-2 Containment (155' elev) Area G-M ( W) Starts IPC dose Perform Steps___________________integration 4.2 R-3 Radiochemistry Lab (AB Area G-M (_W ) Perform Step____139')

____4.3 R-4 #3 Charging Pump (AB Area G-M ( __W) Perform Step__ _ _ 10(Y) __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 4.4 R-5* Spent Fuel Pool Room (AB Area G-M (_.W) Perform Steps__ _ _ 155')_ _ _ _ _ _ _ _ _ 4.5 R-6 Sampling Room (AB 139') Area G-M ( W) Perform Step____ __ ___ __ _ ___ ___ ___ 4.6 R-7 In-core NIS Area (CTMT Area G-M ( W) Starts IPC dose Perform Steps____129', near Seal Table) ____integration 4.7 ,,,,,,,,,,,,,R-8 Drumming Station (AB Area G-M ( W) Perform Step 155')___ ______ 4.8 R-9 SG Sample Panel (Unitl11 Area G-M ( W) No input to Panel) (AB 139') ______this alarm R-10 Penetration Room Filtration APD Beta Semin Perform Step____Discharge (AB 155') ____**(GA-.ES) 4.9 R-I 1

Containment Atmosphere APD Beta Setin Perform Step (AB 121') ____**(GA-ES)

_______4.10

Technical Specification relate-d**Genera Atomics Electronic Systems Page 5 of 15 Vrin7.Version 72.1 V7~Page 1 of 2 07/16/15 08:21:00 [ NP-I -ARP-L1S SETPOINT:

I. High: 153'10" 2. Low: 153'4"I ORIGIN: Level Switch (N1G3 1LS65 1-N)PROBABLE CAUSE LOCATION EH2 SFP LVL 1-I-LO HIGH LEVEL LOW LEVEL Improper valve alignment Normal evaporation

,,Leakage into Spent Fuel Pit TImproper valve ali nent AUTOMATIC ACTION NONE OPERATOR ACTIONS 1. Dispatch operator to determine actual level from the local indicator.

El 2. IF level is low, THE_.__N make up to the spent fuel pit in accordance with FNP-1-SOP-54.0, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM. El CAUTION: SFP pump suction vortexing may occur below 153' 4" level.: (NEL-,98-0329)

E-l1 3. IF level is low AN... evaporation is NOT suspected to be the cause of the low level alarm, THEN perform the following:

a. Check the WHT, FDT, RHTs and RWST for abnormal level increases to determine possible leakage flowpaths.

El b. Check for leakage past the weir gate into the Fuel Transfer Canal or Cask Wash Area. El c. Check for leakage into CTMT by checking Refueling Cavity level (if accessible) and CTMT Waste Sump Level. El d. Perform a SFP liner leak check in accordance with FNP-lI-SOP-5 4.0, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM. El e. Refer To Technical Specifications LCO 3.7.13. El 4. IF level is high, THEN perform the following:

a. Check for any signs of obvious in-leakage to the SFP at the SFP area (e.g. fire protection, demineralized water). El b. Check for other sources of in-leakage to the SFP, if necessary (e.g. Component Cooling Water System, Reactor Makeup Water via the SEP demineralizer).

El c. Terminate source of in-leakage.

El d. Contact Chemistry to obtain current SEP boron concentration.

[I]e. IF boron concentration determined to be <2000 ppm, THEN suspend any movement of fuel assemblies in the storage pool and initiate action to restore the SFP boron concentration to 2000 ppm. (Ref. Technical Specifications LCO 3.7.14) [1 5. Verify proper system valve alignment.

I]6. Return the Spent Fuel Pit Level to normal. El

References:

A-177 100, Sh. 267; D-175043; OCR-383; CN-BC2200; Technical Specifications Page I of I Page of 1Version 59.1 07/1 6/15 08:2 1:50 I t I w7 Page 2 of 2 FNP-2-ARP-1I.5 LOCATION EH2 K SFP LVL HI-LO SETPOINT: 2. ow154" ORIGIN: Level Switch (N2G31LS651-N)

PROBABLE CAUSE HIGH LEVEL LOW LEVEL Improper valve alignment Normal evaporation Leakage into Spent Fuel Pit Improper valve alignent AUTOMATIC ACTION NONE OPERATOR ACTIONS 1. Dispatch operator to determine actual level from the local indicator.

2. IF level is low, THEN make up to the spent fuel pit in accordance with FNP-2-SOP-54.0, spent fuel pit cooling and purification system. (CMT 0003671i CAUTION: SIP pupsuction vortexing may occur below 15Y-4" level. (NEL-,98-0329) 0'-INOTE: The following step does not apply if the low SFP level is expected for an evolution in progress such as SFP feed and bleed. [-'lI 3. I.E level is low AND evaporation is suspected to be the cause of the low level alarm, THIEN~ perform the following:

{CMT 0003671)a. Cheek the WHT, FDT, RilT's and RWST for abnormal level increases to determine possible leakage tlowpaths.

b. Check for leakage past the weir gate into the Fuel Transfer Canal or Cask Wash Area.c. Cheek for leakage into CTMT by checking Refueling Cavity level (if accessible) and CTMT Waste Sump Level.d. Perform a SFP liner leak check in accordance with FNP-2-SOP-54.0, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM.e. Refer To Technical Specification 3.7.13 for LCO Requirements.

El El El El Page I of 2 Page of2Version 50.0 V8 Page 1 of 4 FNP-FSAR-1 2 TABLE 12.1-9 AREA MONITOR ALARM SETPOINTS Channel R-1A (Unit I only)R-1B (Unit 2 only)SR-2 AreaMonitor Control room Technical support center Containment I R-3 R-4 Radiochemistry laboratory Charging pump room 0.75 x 1.0.75 x 10.3 At power 90 x 10"3;after shutdown 20 x 10-3 2.0xl10 3 Inside room 50 x 1 0-2.0 x10"3 15.0Ox 10-3 50.0 x 10-3 15.0 x 10"3 15.0 x 10"3 50.0l lSR-5 R-6 R-7 R-8 R-9 (Unit 2 only)R-27A and B Spent-fuel buildingI Sampling room Incore Instrumentation area Drumming station Sample panel room Contanmenthigh adiaIo a. These s~etpoints are typical of those anticipated during initial plant operation and are subject to change dudng the life of the plant. Actual setlpoints are incorporated in plant procedures.

REV 21 5/08 V8 FNP Units I & 2 RADIATION MONITORING SYSTEM A-1810l5Pae2o4 annunciator on the main control board. The other switch shall actuate on increasing radiation to initiate the RMS High Radiation annunciator on the mai control board. In addition, each function shall actuate an indicating light on the ratemeter front panel defining the alarm condition (References 6.4.034 and 6.4.249).

The ratemeter operation selector switch shall actuate the RMS CH Test annunciator on the main control board when placed in any position other than OPERATE (References 6.4.034 and 6.4.249).3.1.2.3..2 The monitor shall provide an output signal proportional to the indicated radiation level that will be used to provide an isolated signal to the RMDA computer and a recorder (References 6.4.109, 6.7.079 and Open Item Observation RMSoFSD-014).

3.1.1.3.3 The alarm setpoint is based on operating experience (References 6.7.080 and 6.7.096).3.1.2.4 Interface Requirements The instrument power supply for the RMS system panel NIHl INORM 2502A, B, and C is 120 VAC distribution panel lB, breaker number 2 (Reference 6.4.219).

The control power supply for the RMS system panel N1HI INGRM 2502A, B, and C is 2081120 VAC control power panel lN, breaker number 6 (Reference 6.4.107).

The instrument power supply for the RMS system panel N2H1 1NGRM 2502A, B, and C is 120 VAC distribution panel 2B, breaker number 2 (Reference 6.4.345).The control power supply for the RMS system panel N2HI INGR.Ml 2502A, B, and C is 2081120 VAC control power panel 2N, breaker number 6 (Reference 6.4.106).3.1.3 Containment Elevation 155'-4" Area Monitor TPNS No.ND21iRE 0002 3.1.3.1 Basic Function This detector provides a means of monitoring of the containment area to comply with GDC 64 (Reference 6.7.084).23l1U 3431A-I!aI01 3M3- ev 3-4 Rev. 0 V8 FNP Units ! & 2 RADIATION MONITORING SYSTEM A-181015 Page 3 of 4 3.1.6 Fuel Storage Pool Area Monitor TPNS No.ND21IRE 0005 3.1.6.1 Basic Function This detector monitors the spent fuel pool area to comply with the area monitoring criterion of GDCs 63 and 64. A high reading by this monitor is indicative of a loss of shield water in the spent fuel pool (Reference 6.7.084).3.1.6.2 Functional Requirements The monitors shall provide continuous indication over a range of 10.4 to 10' rads per hour (R/hr). In addition, the monitor provides a narrow range indication over a range of 10.4 to 10"' rads per hour (R/hr). The narrow range indication is considered a design feature of the monitors and not a requirement.

The monitors shall provide a flat (:+/-20 percent)response for gamma energies between 100 keV and 2.5 MeV (Reference 6.4.092).3.1.6.3 I&C Requirements 3.1.6.3.1 The monitor ratemeter shall be located in the RMS system panel in the main control room and shall provide a minimum of two single pole double throw alarm relays for external use. Each relay shall be fully adjustable over the entire indicated range of the monitor. One switch shall actuate on a low signal to actuate the RMS Channel Failure annunciator on the main control board. The other switch shall actuate on increasing radiation to initiate the RMS High Radiation annunciator on the main control board. In addition, each function shall actuate an indicating light on the ratemeter front panel defining the alarm condition (References 6.4.034 and 6.4.249).

The ratemeter operation selector switch shall actuate the RMS CH Test annunciator on the main control board when placed in any position other than OPERATE (References 6.4.034 and 6.4.249).3.1.6.3.2 The monitor shall provide an output signal proportional to the indicated radiation level that will be used to provide an isolated signal to the RMDA computer and a recorder (References 6.4.109, 6.7.079, and Open Item Observation RMS-FSD-01 4).~JS6~ )43¶~.tS~OIS RN 3-9 Rev. 0 3-9 Rev. 0 V8 FNP Units 1 & 2 RADIATION MONITORING SYSTEM A-181015 Page 4 of 4 The control power supply for the RMS system panel N2HI INGRM 2502A, B, and C is 2081120 VAC control power panel 2N, breaker number 6 (Reference 6.4.106).3.1.11 Containment High Range Area Monitor TPNS Nos.QD21IRE 0027A, B 3.1.11.1 Basic Function The monitors provide post-accident area monitoring for the containment building in compliance with NUREG 0578, NUREG 0737, and RG 1.97 (Refcrences 6.4.051, 6.4.350, 6.7.001, and 6.7.005).3.1.11.2 Functional Requirements 3.1.11.2.1 The monitor ratemeter shall be located in the RMS system panel in the main control room and shall provide continuous indication over a range of I to 1 0" rads per hour (R/hr). The monitors shall respond to gamma energies as low as 60 keV and provide an essentially flat response for gamma energies between 100 keV and 3 MeY (References 6.4.051, 6.4.350, 6.4.355, 6.7.001, 6.7.003, and 6.7.005).NUREG 0578 originally required a range of I to 10' R/hr in reference to a draft revision to RG 1.97, but this range was lowered tol to l0 7 Rihr by NUREG 0737, Clarification Item II.F.I, Attachment 3 and was subsequently incorporated into RO 1.97, Revision 2. The range selection provides an integrated indication of two RO 1.97 parameters, Type C Category 3 and Type E, Category I, as permitted by RG 1.97, Revision 3, Table 3, Footnote 1.3.1.11.2.2 The monitor shall provide indication in the control room for post-accident monitoring to comply with RG 1.97 (References 6.4.051 and 6.4.3 50).3.1.11.2.3 The monitor shall provide an output signal proportional to the indicated radiation level that may be used to provide an isolated signal to the RMDA computer and/or a recorder.At least one channel must be recorded to comply with the RCG 1.97 PAMd requirements (References 6.4.051, 6.4.109, 6.4.3 50, 6.7.079, and Open Item Observation RMS-FSD-014).U !

I I l015,NU 3-16 3-16 Rev. 0 V9 Page l of 5 FIGURE 5-6 REACTOR VESSEL LEVEL MIMIC DISPLAY 5-36 V9 Page 2 of 5 Procedure Number Ver UN IT 1 Farley Nuclear Plant Ar FNP-1-OOP-4.3 37.0 Page Number 6/3/2015 11:31:52 MID LOOP OPERATIONS 36 of 78 FIGURE 5 I-V9____ ___ ___ ___ __ ___ ___ ___ ____ ___ ___ ___ ___ ___ ___ ___Page 3 of 5 Procedure Number Ver UNIT I Farley Nuclear Plant FNP-1-UOP-4.3 37.0......Page Number 6/3/2015 11:31 :52 MID LOOP OPERATIONS 37 of 78 FIGURE 6 UNIT 1 LEVEL INDICATOR 1717" Top of PRZR (-71,380 Gal)166'3" Top of StG Tubes (-70,722 Gal)..J w I-U-141' -20% przr Lki (-51 ,427 Gal)1351'1 Top of Vessel 1297" Vessel Flange 129'6" Bottom of PPZR (-43,350 Gal)123'11" Nozzle Dam 122'9" Cold Leg Centerline 121'7" Bottom of Hot Leg 119'1" Top of Core 112'5" Bottom of IL 106'10" Bottom of Core 175 170 165 160-155-150-145-140-135-130-125-120-115-110 -100%(166.4')PRZR LI462 0%(134.5')(--27,248 Gal](161.6)LI 2965B 300" (135.7")160'TYGON HOSE 116'(136.7')I130"!RCS SLEVEL LI 2965A O0" V9 Page 4 of 5 Mid Loop Operations FNP-2-UOP-4.3 FARLEY Version 40.0 Unit 2 Page 39 of 83 Page 1 of 1 RCS ELEVATION FIGURE a-IIIIIIIII I I i~runsa uwiw~ui~ a iu:w:uu V9 Pagte 5 of 5...Mid Loop Operations

...........

FNP-2-UOP-4.3 I FARLEY Version 40.0 I Unit 2 Page 40 of 83 FIGURE 6 Page 1 of I RCS LEVEL INDICATION 1001'1 Steam Nozzie Inlet UNT2 IO LEVELI INID ...ICATIION 175 171'7' Topof PRZR 18663 Top of S/G Tubes (-70,722 Gal)w U-_141' -20% Przr Lvi (-51,427 Gall 1351'1 Top of Vessel 1209'7 Vessel Flange 128'8" Bottom of PRZR (.-43,350 G,.I)123'11" Nozzle Dam 12279" Cold Leg Centerline 121"7" Bottom of Hot Leg 110'1* Top of Core 170*160, 155, 150'145'140 135 130 125 120 115 100%(16.4")PlZm LI 462 0%(134.5')(160")LI 2965 150'TYGON HOSE 116'124]121'6"I[30]JRCSI IFTI 1 LEVEl 1 Li w 1125'" Bottom of IL 110 1OG'10" Bottom of Core v.nnheal uwuidzu1 at 14:1:Uuo Vl0o Southern Nuclear Design Calculation Page 1 of 1 SPlant: Farley Unit: 1&2 Calculation Number: SM-SNC524602"001 Isheet: D-1 Attachment D -Water Level Elevations Corresponding to Fuel Uncovery Several EALs are based on the water level elevations that correspond to uncovery of irradiated fuel in the reactor vessel, the spent fuel pool, and the spent fuel transfer canal during refueling operations.

RPV Level at 6" Below Bottom ID of RCS Loop Hot Leg Centerline Elevation

= 122'-9" = 121'-21" [Design Input #9]-1/2A x ID = -'A x 29" = -0'-14.5" [Design Input #9]Elevation

@ bottom of loop = 121'-6.5" (= 121'-21" -0'-14.5")6-inch level reduction

=-0'6 Elevation

@ 6" below bottom of loop = 121'-0.5" = -121'Cold Leg Centerline Elevation

= 122'-9" = 121'-21" [Design Input #91-1/2 x ID = 1/2 x 27.5" --0'-1 3.75" [Design Input #9]Elevation

@ bottom of loop = 121 '-7.25" 6-inch level reduction

=-0'6 Elevation

@ 6" below bottom of loop = 1 20'-1 .25" = -121'Top of Active Fuel (TOAF) in Reactor Vessel The elevation of the top of the irradiated fuel in the reactor vessel is determined by referencing reactor vessel dimensions from drawings U168878 & U206587 to the I.~centerline of the cold and hot leg nozzle centerline elevation

__CL&., as shown to the right. -EL 1229" The upper core plate elevation is the CL/HL centerline

=elevation (122'-9";

Design Input #9) plus the distance to the __, reactor vessel mating surface (82.437";

Design Input #9) then upP~corePat minus the distance from the mating surface to the upper core plate (124.687";

Design Input #9): ELucP = 122.75 ft + [(82.437 in -124.687 in) x (1 ft/12 in)]ELucP = 122.75 ft -3.53 ft ELucP = -119 ft The elevation of TOAF (ELTOAF) is approximately 1 foot below this elevation (Assumption

10): ELTOAF =-118' = 119' -1' Vll Page 1 of 2 FNP-FSAR-1 1 capacity is incorporated in the recycle portion of the LWPS to accommodate abnormal operations.

The basic composition of the liquid collected in the recycle holdup tank is boric acid and water with some radioactivity.

Liquid collected in this tank is evaporated to remove radioisotopes, boron, and air from the water so that it may be reused in the reactor coolant system.Evaporator bottoms are normally processed at a low boron concentration to the waste holdup tank unless found acceptable for boric acid recycle. The condensate leaving the waste evaporator may pass through the waste condensate demineralizer and then enter the condensate tank. When a sufficient quantity of water has collected in the waste condensate tank, it is normally transferred to the reactor makeup water storage tank for reuse. Samples are taken at sufficiently frequent intervals to ensure proper operation of the system to minimize the need for reprocessing.

If a sample indicates that further processing is required, the condensate may be passed through the waste condensate demineralizer or, if necessary, returned to the recycle holdup tank for additional evaporation.

The water collected in the recycle holdup tank may be routed to the disposable demineralizer for processing rather than processing the water through the evaporator.

Water processed through the demineralizer is not normally recycled.11.2.2.2 Waste Portion (Drain Channel B -Nonreactor Grade Water Sources)Drain channel B is provided to collect and process nonreactor grade liquid wastes. These include floor drains, equipment drains containing nonreactor grade water, laundry and hot shower drains, and other nonreactor grade sources. Drain channel B equipment includes a floor drain tank and filter, laundry and hot shower tank and filter, chemical drain tank, waste monitor tank demineralizer and filter, disposable demineralizer system, and two waste monitor tanks.Nonrecyclable reactor coolant leakage enters the waste holdup tank from system leaks inside the containment via the containment sump and enters the floor drain tank from system leaks in the auxiliary building via the floor drains. Unless an extremely large leak develops, this liquid woul not e recycled ecauseit is diluted and contaminated ywaterentening te foor drain tank from other sources, e.g., laboratory equipment rinses, hose water, component cooling leaks, etc. Nonreactor grade leakage enters the floor drain tank from the auxiliary building floor drains. Sources of water to the drains are fan cooler leaks, secondary side steam and feedwater leaks, component cooling water, and hose water. This leakage is assumed not to contribute significantly to activity release. The activity level is normally much less than Normally, the activity of the floor drain tank contents is well below permissible levels. Hence the contents may be transferred directly to the waste monitor tanks after sampling.

Following analysis to confirm the acceptable low level, the tank contents are discharged without further treatment.

However, should spills, leaks, or equipment failures cause radioactive water to enter the floor drain tank, this water is processed through the waste evaporator or disposable demineralizer.

11.2-3 11.2-3REV 23 5/11 V11 Page 2 of 2 FNP-FSAR-11I G. Floor Drain Tank Pump One standard pump is used to transfer water normally to the waste monitor tank.The pump can also be used to supply the waste evaporator or for pumping the waste back to the waste holdup tank.H. Waste Monitor Tank Pumps One standard pump Is used for each tank to discharge water or to recycle water if further processing is required.

The pump may also be used for circulating the water in the waste monitor tank in order to obtain uniform tank contents and hence a representative sample before discharge.

The pump can be throttled to achieve the desired flowrate.11.2.3.1.2 Reactor Coolant Drain Tank Heat Exchangers The reactor coolant drain tank heat exchanger is a U-tube type with one shell pass and two tube passes. Although the heat exchanger is normally used in conjunction with the reactor coolant drain tank, it can also cool the pressurizer relief tank from 200°F to 120"F in < 8 hi.11.2.3.1.3 Tanks A. Reactor Coolant Drain Tank One tank is provided for each unit. The purpose of the reactor coolant drain tank is to collect leakoff type drains inside the containment at a central collection point for further disposition through a single containment penetration via the reactor coolant drain tank pumps. The tank provides surge and net positive suction head requirements to the pumps.The water entering the reactor coolant drain tank may be of adequate purity to allow direct recycling to the boron recycle system holdup tank. If this water is not compatible or if it contains dissolved air or nitrogen, it must be processed in the LWPS channel A.Sources of water entering the reactor coolant drain tank include the reactor vessel flange leakoff, valve leakoffs, reactor coolant pump 2 and 3 seal leakoffs, and the excess letdown heat exchanger flow. No continuous leakage is expected from the reactor vessel flange during operation.

The system is designed to maintain a constant level in the tank to minimize the amount of gas sent to the waste gas processing system and also to minimize the amount of hydrogen required.

One pump runs continuously.

The level in the tank is maintained by a control valve in the discharge line. The valve operates on signals from a level controller connected to the tank and regulates flow fractions back to the tank and out of the system, respectively.

11.2-6 11.2-8REV 23 5/111 V12, Page 1 of 4 Southern Nuclear Operating Company UOt~l lnit: FNP SM-SNCSheet 00 I t Pl a nit: FNP Title: NEI 99-01 Rev 6 EAL Calculations SCheet -0 38 CG1 Loss of RPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

Emergency Action Level: Cold Shutdown Refueling (1)1.a. RVLIS (Mode 5) level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following:

Containment High Range Radiation Monitor RE27A OR 27B reading >100 R/hr.* Erratic source range monitor indication.

UNPLANNED rise in Containment Sump, or Ractor Coolant Drain Tank (RCDT), or Waste Holdup Tank (WHT) levels of sufficient magnitude to indicate core uncover.The Containment High Range Radiation Monitor reading corresponds to the reflected dose rate from the irradiated fuel in the RPV with an RPV water level of<_ EL 1 18'O", TOAF. It is calculated in Attachment F3 of this calculation.

AND c. ANY indication from the Containment Challenge Table C1 Containment Challenge Table Cl CONTAINMENT CLOSURE not established*

> 6% H2 exists inside containment UNPLANNED increase in containment pressure*If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Sheets 9 and 12 of FNP SAMG calculation SM-95-O754-OOlestablish the 6%by volume hydrogen limit.

V12, Page 2of 4 Southern Nuclear Operating Company I Plant: FNP ITISM-S NC524602-001 OMAV Unit: 1& Title: NEI 99-01 Rev 6 EAL Calculations Sheet 4 to one significant digit (page 15 of U26469 8), thus the value of 8,000 P/hr is appropriate.

The calculation used reactor coolant system fission product concentrations from WCAP14 722 page 7.6-52 to calculate isotopic concentrations.

The calculation for DEI131 was performed to find a ratio to DEl 0.5 pCi/gm. GRQDEC was used for energy binning, a geometric factor and dose conversion factors taken from SM-94-0466-001 was applied.Containment Barrier Potential Loss Threshold 4.8 Containment Hydrogen Concentration greater than 5.5%.Per calculation SM-95-0754-001 pg 9, a hydrogen concentration above 6% is potentially explosive.

Since the accuracy of the hydrogen monitor is +/-0.5% in the range of 0-10%hydrogen, use 5.5%. Per calculation SM-95-0754-001 page 12 the concentration of>6% would support a burn throughout containment.

V12 Page 3 of 4 Calculation Sheet Southern Company Services g rrojcc Farley Nuclear Plant CalculatoaNunc SM-95-0754-001 ISevere Accident Management Guidelines I 9 of 147 1 Parameter/

SAMC Sctpolnt No.: Containment Hydrogen / H01 Set point Value

Description:

Upper limit of operability of the hydrogen recombiner

[Detemine Setpint alue6 % by volume ( See NOTE)]Specific Setpoint Usage: This selpoint is used within SAG-7 (REDUCE CONTAINMENT HYDROGEN) to determine if using the recombiner may jeopardize its future use.Specific AR -2703A (1-12/O02 analyzer train "A"); AR -2703B (112/02 analyzer train (idnzSAMG Technical Basis Section 6.0, SAG-7 Referenes:~

FNP Vendor Manuals # UJ-21 232C and U-262920, VNP Vendor Manual #1/2AE01-38-2, FNP FSAR 622.5 & 7.64 and VNP FSAR. WCAP-7709-L, Supplement 5; P&ID Nos. D-175019 & D-205019~This setpoint is the containment hydrogen conceniration at which the recombiner may overheat, based on information in the hydrogen recombiner systems manual.NOTE: A review of available information in the FNP H, recombiner system manual was conducted.

Other than the recommended H, concentration design limit of 4% by volume, no upper limit hydrogen concentration value was noted in FNP-112 system manual U-212326C.

The FSAR Past-LOCA H, production analysis and recombiner performance analysis do not discuss or identify any specific H, concentration upper limits relative to recomnbiner saturation or overheating.

The FNqP PSAR referenced WCAP-78'20('Non-Proprietary) to address prototype testing. WCAP -7709-L, the proprietary version of WCAP-7820, was reviewed and it indicated that the recombiners had been successfully performance tested to 6.2 % by volume of H 2 in dry air. This value is conservative when an accident/steam environment is considered.

It was also tested to demonstrate operation at up to 77 psia in a steam environment.

A review of the FNP 11,1O, analyzer design and system manual indicates that the measured concentration values are on a wet H, measurement basis since sample lines are heat traced and the analyzer has a pre-analysis "hot box" area@ 275 ° to prevent any steam condensation in the sample which could change the abundance of measured gas. Furthermore, the indicated range for the H, /02 analyzer is 0-10 volume % H 2 .I iuaku2Per the vendor manuals, the accuracy of the analyzer is 5% or the equivalent of 10.5% H 2.Thus the setpoint is rounded to 6%.8t V12 Page 4 of 4 Calculation Sheet Southern Company Services 1 talkie .Y^. ...U i'roject Farley Nuclear Plant I Lailcatlon Numbnlr SM.95-40754-00 1 ll SubjectilTil I Sheet Severe Accident Management Guidelines 12 of 147 Parameter/

SAMG Setpoint No.: Containment Hydrogen / H-03 Set point / Value Dlescription:

Hydrogen concentration that may result in a burn if the hydrogen recombiner is used.iDetermined Setpoiot Value: > 6 % by volume (SeeNOTE)I Specific SetpointUsage:

This setpoint is used within SAG-7 (REDUCE CONTAINMENT HYDROGEN)to determine the use of the recombiner that may cause a hydrogen burn.Specific Instrumentts):

AR -2703A (H 2 /02 analyzer train "A"); AR.- 2703B (H-2 /02 analyzer train adlmiSAMG Technical Basis Section 6.0, SAG-7 Reeecs FNP Vendor Manuals # U-21232C and U-262920, VNP Vendor Manual #l/2AEOI-38-2, FNP FSAR 6.2.5 & 7.6.4 and VNP ESAR, WCAP -7709 -L, Supplement 5; P&ID Nos. D-175019 & ID-205019.Asumptons:This setpoint is the containment hydrogcn conccuntration at which the recombiner may cause a global H 2 burn, based on test infornmationi in WCAP -7709-L.NOTE: Like setpoint HO1, a review of available inforrmationi in the FNP H 2 recombiner system manual and other documentation was conducted.

No global burn limit H 2 concentration value was noted in FNP-l,2 system manual U-212326C or other references.

The FSAR Post-LOCA H" 2 production analysis and recombiner performance analysis do not discuss or identify any specific 112 concentration upper limits relative to recombiner saturation or overheating.

The FNP FSAR referenced WCAP-Tg20(Non-Proprietary) to address prototype testing. WCAP-7709-L, the proprietary version of WCAP-7820, was reviewed and it indicated that the recornbiners had been successfully performance tested to 6.2 %A by volume of H 2 in dry air. This value is conservative when an accidentlsteamn enviroinment is considered.

It was also tested to demonstrate operation at up to 77 psia in a steam environment.(asaznnPer the vendor manuals, the accuracy of the analyzer is 5% or the equivalent of 0.5% H 2.Thus the setpoint is rounded to 6%.I S1 w V1 3 Page 1 of 6 V1 3 Page 2 of 6 V1 3 Page 3 of 6 V1 3 Page 4 of 6 V1 3 Page 5 of 6 V1 3 Page 6 of 6 V14 Page 1 of 2 A-I 81004 FNP Units 1 & 2 ELEC DIST SYSTEM E3.3.1 Basic Functions E3.3.1.1 Switching

-The 125 V dc switchgears shall provide circuit switching and isolation capabilities between the battery, battery chargers, dc distribution panels, inverters, and diesel generator control panels during all plant operating modes, as required by plant operation.

E3.3.1.2 Protection

-The 125 V dc switchgear circuit breakers shall provide selective tripping of circuits for overload or short circuit conditions to isolate the failed feeder and to minimize system disturbance (References E6. 1.001, E6.3.002, E6.4.0 19, E6.4.020, E6.4.023, E6.4.024, E6.5.003).

Functional Requirements E3.3.2 E3.3.2.1L E3.3.2.2 125 V dc switchgears nominally rated at 125 V dc shall be suitable for continuous operation within a range of[ 105 to 140 V (Reference E6.5.003).

125 V dc switchgear bus ratings shall be:* Bus rated continuous current -1,000 A E3-14 E3-14Ver.

44.0 I V1 4 Page 2 of 2 FN P-0-EMP-1 340.10 07/09/15 08:03:35 TABLE 1 Table 1 Auxiliary Building Battery -Approximate Remaining Hours Battery Discharge Remaining Hours Starting from Battery Terminal Voltage Below Amps (Rate) 117 Volts 114 Volts 111 Volts 100 Amps 12 h 6 h 2 h 150 Amps 9 h 4 h 1.3 h 200 Amps 7.5 h 3.8 h 1.3 h 250OAmps 7 h 3 h 1 h 300 Amps 6 h 2.8 h 1 h 350 Amps 5 h 2.6 h 1 h Remaininn JNote: When the battery terminal voltage reaches 15Vj the battery is fully discharged.

Assumptions:

1. It is assumed that the average battery electrolyte temperature is 77°F. For temperatures higher than 77, the capacity increases and therefore the estimated time will be lower than actual. However, for temperatures lower than 77, the capacity decreases and the estimated time will be higher than actual.2. It is assumed that the battery capacity is at 100% of rated capacity based on the latest performance test done on the battery (No aging correction).

The estimated time is directly proportional to the capacity but the relationship is not linear.3. For conservatism, it is assumed that the design margin used in sizing is zero.4. It is assumed that the battery consists of 60 connected cells. If one or more cells are bypassed, the actual time will be less than that estimated.

Page 1 of 1Veso20 Version 2.0 Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant Units 1 and 2;Edwin I. Hatch Nuclear Plant Units 1 and 2;Vogtle Electric Generating Plant Units 1 and 2;License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 4 EAL Verification and Validation Documents Joseph M. Farley Nuclear Plant Units 1 and 2 License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Enclosure 4 Farley EAL Verification and Validation Documents Farley EAL V&V Documentation Index V&V# ~Descripjtion -IC i Threshold; RCS Modes/TemPeratures

-TS Reference (Table 1.1-1) EAL Basis Section 1.4 1 RCS Cold Shutdown Temperature reference (200 °F) CA3 (1)CU3 (1)Radiation Monitor Calculations:

RG 1 (1)RE-15C RS1 (1)RE-29B (NG) RA1 (1)Calc SM-SNC524602-O01 Radiation Monitor Calculations:

RUl1 (1)RE-18 RE-23B RE-15 R- 14 RE-22 RE-29B 2 Caic SM-SNC52 460 2-OO1 Radiation Monitor Calculations:

CG 1 (1)b RE-27A&B CS1 (2)b Caic SM-SNC524602-O01 Containment Radiation Monitor Calculations:

FCB L3A FCB L3A RE-27A&B CB PL3A CB PL3A Caic SM-SNC52 4602-O01 Containment Radiation Monitor Calculations:

RCB L3A RCB L3A RE-2 RE-7 Calc SM-SNC52 4602-O01 ODCM/TS Reference to Site Boundary RG1 (2)(3)3RS1 (.2)(3)RA1 (2)(3)(4)SFP Level 3 & 2 Indications RG2 (1)4 RS2 (1)RA2 (3)Alarm Response -FNP- 1(2)-ARP-1.6 RA2 (2)5 Annunciator Display: RE-24A&B RE-5 RE-25A&B____

Alarm Response -FNP- 1-ARP- 1.6 Control Room Rad RA3 (1)6 Monitor (Annunciator Display RE-l1A)______

Alarm Response -FNP-I(2)-ARP-I

.5 (SFP Level) RU2 (1)a Annunciator Display: EH2 (SFP LVL HI/LO)Radiation Monitor Information (FSAR): RE-5 RU2 (1)b 8 RE-2 RE-27A&B 9 RVLIS/RPV Level CS 1 (l)b FNP-1(2)-UOP-4.3 CAl (1)10 RPV Level Calculation CS1 (1)b Containment Sump/Reactor Coolant Drain Tank CG 1 (1)b 11 (RCDT)/Waste Holdup Tank (WHT) FSAR Reference CS 1 (2)b CAl (2)b CU1 (2)b H 2 Concentration

(> 6%) -calculation/reference CG1 (1)c 12 Calc SM-95-0754-001 CB PL4B CB PL4B 2/22/2016 Page 1 of 2 Farley EAL V&V Documentation Index V&V# Description IC Th reshold ESF Busses Drawing CA2 (I)CU2 (1)a SG1 (1)a 13 SG2 (1)a SS1 (1)SA1 (1)SU1 (1)DC Voltage Reference CU4 (1)14 SG8 (1)b SS8 (1)15 ISFSI TS/Dose Reading Calculation E-HU1 (1)Caic SM-SNC5 2460 2-O01 CSFST Information FCB PL1A FCB PL1A RCB PL1B RCB PL1B FCB L2A FCB L2A FCB PL2A FCB PL2A FCB PL2B FCB PL2B 16 RCB PL2A RCB PL2A CB PL2A CB PL2A CB PL4A CB PL4A CB PL4C. 1 CB PL4C. 1 SG1 (1)b 9S5 (1)c Rad Monitor Reference -FNP-I (2)-ARP-1.6 CB L4B CB L4B RE- 10 17 RE-14 RE-21 RE-22 Seismic ARP -FNP-l(2)-ARP-1.12 HU2 (1)18 (Annunciator Display (K5))______

RCS sample activity -Tech Spec Values SU3 (1)19 TS 3.4.16 RCS Leakage -Tech Spec Values SU4 (1)20 TS 3.4.13 ______ (2)21 Containment Spray Initiation Setpoint (27 psig) SU7 (2)a 2/22/2016 Page 2 of 2 Vl Page 1 of 1 Definitions 1.1 Table 1.1-1 (page 1 of 1)MODES MODE TITLE REACTIVITY

% RATED AVERAGE CONDITION THERMAL REACTOR COOLANT (keff) POWER(a) TEMPERATURE (0 F)1 Power Operation

_> 0.99 > 5 NA 2 Startup >_0.99 _<5 NA 3 Hot Standby < 0.99 NA _> 350 4 Hot Shutdown(b)

< 0.99 NA 350 > Tavg > 200 5 Cold Shutdown(b)

< 0.99 NA _<200 6 Refueling(C)

NA NA NA (a)(b)(c)Excluding decay heat.All reactor vessel head closure bolts fully tensioned.

One or more reactor vessel head closure bolts less than fully tensioned.

Farley Units 1 and 2 1.1-7 Amendment No. 146 (Unit 1)Amendment No. 137 (Unit 2) 7 V2 See NL-1 5-1 898 Enclosure 3 SIUTrHEiRI A, Southern Nuclear Design Calculation-Calculation Number: SM-SNC524602-001 Plant: IFarley Nuclear Plant Un:171-2 ]1& Dsilr"M hicl Title:

Subject:

Emergency NEI 99-0 1 Rev 6 EAt Calculations IAction Level Setpoints Purpose I Objective:

Document Emergency Action Level Values to support conversion to NEI 99-01 Rev 6 SSystem or Equipment Tag Numbers: NlA Contents Topic Page Attachments

of (Computer Printouts, Technical Papers, Pages_________________Sketches, Correspondence),Purpose 1 A-"SNC EP Concurrence 1 Criteria 1 B -Reserved 0 Conclusions