ML19296C608

Forwards Response to NUREG-0578 Re Lessons Learned Task Force short-term Requirements as Requested by NRC .Twelve Oversize Drawings Encl

ML19296C608
Person / Time
Site:	Farley
Issue date:	02/21/1980
From:	Clayton F ALABAMA POWER CO.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8002260702
Download: ML19296C608 (68)

Text

{{#Wiki_filter:'* Alabama Power Company a 600 Nortn 18th Street Post office Box 2641 Birmingnam. Alabama 35291 Telephone 205 323-5341 L leo"'fc'e',1Sni Alabama Power vaunwn w rc wem Febrinry 21, 1980 Docket No. 50-364 Mr. Domenic B. Vassallo, Acting Director Division of Project bunagement U. S. Nuclear Pegulatory Ccrmission Washington, D. C. 20555

Dear Mr. Vassallo:

As requested by your letter of September 27, 1979, Alabama Power Capany sutmits Enclosure (1) documenting ccmnitments to hTREG-0578 requirements and January 1, 1981 requirements, which are addressed with appropriate conceptual designs. As additional infomation is supplied by the Division of Project Ihnage-ment regarding requirements in the areas of Iassons Icarned or as further study by Alabama Power Company requires, the ccmnitments mntained in Enclosure (1) will be amended. Yours very truly, ,s C(i ,G.t 3 F. L. Clayton, Jr. -l FICjr/INE/nnb Enclosure Peferences: (1) NUREG-0578, "'IMI-2 Iassons Learned Task Force Status Peport and Short Tem Pecor:mendations", July, 1979. (45 cys.) (2) " Follow Lp Actions Besulting from the NPC Staff Paviews Pegarding the 'Ihree Mile Island thit 2 Accident", September 27, 1979. (3) Handouts at Atlanta Pegional Sbeting, "Pegional Meetings 'IMI Short Term Implementation Action"7 September 28, 1979. \\ 6 cc: Mr. R. A. 'Ihorus %C(6 l (i y Mr. G. F. Trchtridge q Q* \\ 1 hfo$ea0 Vg ga e00e26o CL

JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 RESPONSE TO SHORT-TERM LESSONS LEARNED: NUREG-0578

Farley Nuclear Plant Unit 2 Docket No. 50-364 TABLE OF CONTENTS TITLE PAGE SECTION 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-0perated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs.. 1 SECTION 2.1.2 Performance Testing for BWR and 2 PWR Relief and Safety Valves SECTION 2.1.3.A Direct Position Indication of Relief and Safety Valves. 3 SECTION 2.1.3.B Instrumentation for Detection of Inadequate Core Coolings in PWRs. 4 SECTION 2.1.4 Containment Isolation Provisions for PWRs and BWRs. 8 SECTION 2.1.5.A Dedicated Penetrations for External Recombiner or Post-Accident External Purge System. 10 SECTION 2.1.5.C Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant. 11 SECTIC:! 2.1.5.3 Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRS.. 12 i

Farley Nuclear Plant Unit 2 Occket No. 50-364 TABLE OF CONTENTS (Continued) TITLE PAGE SECTION 2.1.6.P Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations. 13 SECTION 2.1.7.A Automatic Initiation of the Auxiliary Feedwater System. 14 SECTION 2.1.7.B Auxiliary Feedwater Flow Indication 16 to Steam Generators for PWRs.. SECTION 2.1.8.A Post-Accident Sampling Capability 17 ATTACHMENT: Description of Post-Accident Sampling Systems...... 18 SECTION 2.1.8.8 Interim Procedures for Quantifying High Level Accidental Radioactivity Releases 23 SECTION 2.1.8.B1 High Range Effluent Motor 24 SECTION 2.1.8.B2 High Range Effluent Radiciodine and Particulate Sampling and Analysis 25 SECTION 2.1.8.B3 Hign Conta': ment Radiation 26 SECTION 2.1.8.C In-Plant Iodine Instrumentation 27 ii

Farley Nuclear Plant Unit 2 Docket No. 50-364 TABLE OF CONTENTS (Continued) TITLE PAGE SECTION 2.1.9 Transient and Accident Analysis 28 CONTAINMENT PRESSURE INDICATION (ACRS) 29 CCNTAINMENT WATER LEVEL MONITOR (ACRS) 30 31 CONTAINMENT HYDR 0 GEN INDICATION (ACRS) REACTOR COOLANT SYSTEM VENTING 32 SECTION 2.2.1.A Shift Supervisor Responsibilities 34 SECTION 2.2.1.B Shift Technical Advisor 35 SECTION 2.2.1.C Shift and Relief Turnover Procedures.... 36 SECTION 2.2.2.A Control Room Access 37 SECTION 2.2.c.S 38 Onsite Technical Support Center SECTION 2.2.2.C Onsite Operational Support Center 46 iii

Farley Nuclear Plant Unit 2 Docket No. 50-364 LIST OF FIGURES TITLE PAGE Figure 1 Reactor Vessel Level Instrumentation. 7 Figure 2 Reactor Coolant Sampling System 19 Figure 3 Reactor Coolant Post-Accident Sampling Panel 20 Figure 4 Vent Stack Effluent Sampling System 21 22 Figure 5 Containment Air Sampling System Figure 6 Flow Diagram of the Reactor Vessel Head Vent System 33 Figure 7 Control Room Interim Technical Support Center.. 41 Figure 8 Technical 5ipport Center.. 42 Figure 9 Control Room CCTV Monitoring System 43 Figure 10 Technical Support Center Control Room CCTV Monitoring System 44 45 Figure 11 HVAC-Technical Support Center Figure 12 Control Room Operational 47 Support Center. iv

Farley Nuclear Plant Unit 2 Docket No. 50-364 LIST OF DRAWINGS TITLE NUMBER PAGE Auxiliary Feedwater D-205007 50 System Main Steam and D-205033 51 Auxiliary Steam Systems Single Line - Electrical D-207001 52 Auxiliary System (Emergency) 4160 V and 600 V) Single Line - Protection D-207005 53 and Metering 4160 V Switchgear, Bus 2F Single Line - Protection 0-207006 54 and Metering 4160 V Switchgear, Bus 2F Single Line - 120 Vac D-207024 55 Vital and Regulated System A Single Line - 120 Vac D-207025 56 Vital and Regulated System B Block Diagram A-207076 57 Auxiliary Feedwater Pump D-207186 58 41EDV No. 2A & 2B Turbine Driven Auxiliary Feedwater Pump Train "C" (Sheet 1 of 2) D-207188 59 (Sheet 2 of 2) 0-207189 60 Solenoid Valves D-207590 61 Solenoid Valves D-207591 62 Solencio Va hes D-207857 63 v

Earley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs Pressurizer Heater Power Supply Baged on the analysis performed by Westinghouse for a 3-loop PWR with a 1400-ft pressurizer, a minimum of 125 kW pressurizer heater capacity is required to establish and maintain natural circulation following a loss of offsite power. This capacity is also sufficient to maintain natural circulation with small break LOCA and subsequent isolation. Westinghouse recommends that these heaters have the capability to be energized within 1 hour following the loss of power. The Farley Unit 2 pressurizer is equipped with 78 individual heater elements assembled in 5 groups. Heater groups A and B each contain 15 ele-ments with a heater capacity of 269.25 kW per group. These heater groups have the capability of being powered from the emergency section of the 600-V load centers 2A and 2C, respectively. These heater groups can be powered from either the normal or emergency bus by operation of breakers controlled from the control room. The pressurizer heater backup groups A and B are redundant, and their associated circuits meet the separation requirements for redundant systems. Procedures and training are included in the Unit 2 operator training program to make the operator aware of when and how the required pressurizer heaters are to be connected to the emergency buses. The procedures identify under what conditions other emergency loads may be shed from the buses in order to provide suf ficient capacity for connection of the required heaters. Power Supply for Power-0perated Relief Valves (PORV) and Block Valves Farley Nuclear Plant has two PORVs instclied. Their associated circuits meet the separation requirements for redundant systems with one PORV powered from train "A" and the other powered from train "B". Each PORV is air operated and equipped with two solenoid valves in their respective air lines. The sole-noids are powered from train A ano B 125-Vdc buses, respectively. These buses are powered through associated battery chargers which can also be sup-plied from the train A or B diesel generator upon loss of offsite power or from 125-Vdc batteries. Each PORV has a motor-operated block valve located in its respective piping. These valves' associated circuits meet the separation requirements for redun-dant systems. The operator and its associated control circuits for each block valve ar" 'e~s'ed f rom its associated train A or B emergency bus. Pressurizer Level Indication Power Supply There are three channelized pressurizer level indicators that receive signals from their corresponding channelized level transmitter. The power supply for each channelized indicator is provided by the correspondent 120-Vac vital and regulated channel (e.g., emergency bus). 1

Farley Nuclear Plant Unit 2 Docket 50-364 $ECTION 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves By letter dated December 17, 1979, Mr. William J. Cahill, Jr., Chairman of the EPRI Safety and Analysis Task Force submitted the " Program Plan for the Per-formance Verification of PWR Safety / Relief Valves and Systems," December 13, 1979. Alabama Power Company considers the program to be responsive to the require-ments presented in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Shcrt Term Recommendations" dated July 1979, Item 2.1.2 which recommended, in part, to " commit to provide performance serifications by full scale proto-typical testing for all relief and safety valves. Test conditions shall include two phase slug flow and subcooled liquid flow calculated to occur'for design basic transients and accidents." The EPRI Program Plan provides for a L- ' etion of the essential portions of the test program by July 1981. Alabama Power Company will be participating in the EPRI program to provide p,ogram review and to supply plant specific data as required. 2

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.3.A Direct Position Indication of Relief and Safety Valves The pressurizer power-operated relief valves have sten-mounted limit switches which control red and green indicating lights in the valva control switch on the main control board (MCB). This indicating system provides positive open and closed indication for these valses. The presently installed limit switches will be replaced with qualified limit switches and an alarm function associated with the valve position will be added prior to receipt of an operating license. This alarm function will be provided on the main control board. The pressurizer safety valves will be modified to install qualified stem-mounted limit switches which will provide positive open and shut indication and the associated alarm function on the main control board. This modification will be completed prior to receipt of an operating license. As backup indication, there are temperature detectors in all relief and safety lines to the pressurizer relief tank. In addition, temperature, pressure, and level indication for the pressurizer relief tank are provided on the MCB and alarmed in the plant computer. These indications are available to the oper-ator to confirm the position of the pressurizer power-operated relief valves and safety valves. 3

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.3.8 Instrumentation for Detection of Inadequate Core Cooling in PWRs Procedures and Description of Existing Instrumentation The Westinghouse Owners' Group, of which Alabama Power Company is a member, has performed analyses as required by Item 2.1.9 to study the effects of inadequate core cooling. These analyses were provided to the NRC " Bulletins and Orders Task Force" for review on October 31, 1979. As part of the sub-mittal made by the Owners' Group, an " Instruction to Restore Core Cooling during a Small LOCA" was included. This instruction provides the basis for procedure changes and operator training required to recognize the existence of inadequate core coolirg and restore core cooling based on existing instrumen-tation. Alabama Power Company has incorporated the key considerations of this instruction into the Unit 2 operator training program. Subcooling Meter Alabama Power Company will install a primary coolant saturation meter that meets the requirements of NUREG-0578. This saturation meter will be installed prior to receipt of an operating license. Addition Instrumentation to Indicate Inadequate Core Cooling The submittal references in " Procedures and Description of Existing Instru-mentation" above described the capabilities of the core exit thermocouples in determining the existence of inadequate core cooling conditions and their superiority in some instances to the loop RTDs for measuring true core con-ditions. Other means of determining the approach to or existence of inade-quate core cooling could be: 1. Reactor vessel water level. 2. Incure detectors. 3. Excore detectors. 4. Reactor coolant pump motor currents. 5. Steam generator pressure. A discussion of the possible use of these measurements is addressed below. The use of incore movable detectors to determine the existence of inadequate core cooling conditions appears doubtful. The detectors could be driven into the tops of the incore thimbles, which are located at the top of the core, following an accident in which concern for inadequate core cooling exists. The problem comes in the lack of sensitivity of the detectors to very low neutron levo'c and changes that would occur due to core uncovery. Gamma detectors could parhaps be employed, but they suffer from similar sensitivity problems and the fact that gamma levels in the fuel region change insignifi-cantly between the covered and uncovered condition. As a result, it does not appear worthwhile to pursue incore movable detectors as a means of determining inadequate core cooling conditions. A

Farley Nuclear Plant Unit 2 Docket No. 50-364 Section 2.1.3.8 The use of excore detectors has been mentioned as a possibility in responding to core uncovery. The only detectors which would have the required sensi-tivity are the source range monitors, since the intermediate and power range monitors are not sensitive enough to the low level changes resulting from vessel voiding. The use of the source range monitors will be investigated further as part of the more indepth study of inadequate core cooling being performed by the Westinghouse Owners' Group. However, their use is probably limited to those instances when significant voiding exists in the downcomer region, since normally water in the downcomer would effectively shield the detectors from the core region whether voids existed or not. Reactor coolant pump motor current, which could be indicative of core voiding, is inappropriate for a reliable means of determining inadequate core cooling, since a loss of offsite power or pump trip because of a LOCA blowdown shuts the pumps down. Steam generator pressure, which already exists, is useful in the case where heat transfer from primary to secondary is interrupted due to loss of natural circulation. This, however, does not satisfy requirements to indicate the approach to inadequate core cooling, nor does it indicate the true condition of the core. Reaction vessel water level determination is the most promising of the items discussed to provide additional capability of determining the approach to and the existence of inadequate core cooling. Several systems for determining water level are under review by the Westinghouse Owners' Group. A conceptual design of one system is given below. Vessel Level System Description Af ter examining many different methods and principles for determining the water level in the reactor vessel, a basic delta pressure measurement from the Dottom of the vessel to'the top of the vessel appears to provide the most meaningful and reliable information to the operator. One of the reasons for choosing this system is that the sources of potential errors are better known for this system than for any other new or untested system. Figure 1 shows a simplified sketch of the proposed vessel level in-tation system. The bottom tap of the instrument would use a thimb~ the incore movable detector system either at the seal table or in the - ale below the vessel. Use of the thimble as part of the incore flux mcattoring would not be lost. The flux thimble guide tube would be tapped below the vessel and an instrument line connection made. The instrument line would have an isolation valve and slope down to a hydraulic coupler connected to a sealeu r-for ce leg. For connection at the seal table, a special fitting would be u alized which would be connected to an isolation valve and sealed reference leg. The sealed reference leg would go to the differential pressure transmitter located at a higher elevation above the expected level of contain-ment flooding. A similar sealed leg would go to the top of the vessel and penetrate the head using the vent line or a special connection on a spare RCC mechanism penetration. Two trains of vessel level instrumentation would be provided. 5

Farley Nuclear Plant Unit 2 Docket 50-364 Section 2.1.3.8 The behavior of the signal generated by this level instrument under normal and accident conditions is being evaluated. The usefulness of this instrument to provide an unambiguous indication of inadequate core cooling is being evalu-ated as part of Item 2.1.9. The potential errors and accuracy of a final system configuration are being evaluated to assess its usefulness to provide information to the operator for proper operation of a vessel venting system, and for normal water level control during periods when the primary system is open and a water level may exist in the vessel. The connection of the level system to the vessel head should be designed to be compatible with the head vent system. Operation of the vent system should not upset all indications of vessel level. This can easily be avoided by using a separate instrument tap or by using more than one location. 6

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Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.4 Containment Isolation Provisions for PWRs and BWRs There are two phases of containment isolation at Farley. Phase A isolates all penetrations except component cooling water, containment spray, and systems essential for safe shutdown. Phase B isolates all remaining process lines except safety injection, containment spray, service water lines to containment coolers, and auxiliary feedwater. Phase A isolation is initiated by all safety injection signals or manual initiation. Phase B isolation is initiated by containment pressure or manual initiation. A high radiation signal is also used for purge isolation. Resetting of isolation signals will not reopen isolation valves. Manual action is needed to open each valve. In addition, Alabama Power has reviewed containment isolation design aGainst a generic study generated by the Westing-house TMI Owners' Group. Farley containment isolation design meets all re-quirements of this study. A list of the systems identified as essential and nonessential appears below. Essential Systems 1. Normal letdown 2. Excess letdown / seal water return 3. Pressurizer sample 4. Hot leg sample 5. Containment air sample 6. Normal charging 7. Residual heat removal (normal suction) 8. Residual heat removal (containment sump recirculation) 9. High head safety injection 10. Low head safety injection 11. Containment spray 12. Containment spray (containment sump recirculation) 13. Service water to containment coolers 14. Reactor coolant pump (RCP) seal water supply 15. Containment pressure 16. Instrument air supply 17. Component cooling to RCP thermal barrier 18. Post-accident air sample 19. Post-accident containment vent Nonessential Systems 1. Acctnulator test lines 2. Accumulator makeup 3. Accumulator sample 4. Nitrogen to accumulators 5. Nitrogen to pressurizer relief tank (PRT) 6. PRT makeup 8

Farley Nuclear Plant Unit 2 Docket No. 50-364 Section 2.1.4 7. Reactor coolant drain tank (RCDT) drain 8. Containment differential pressure 9. Component cooling to excess letdown and RCDT heat exchanger 10. Component cooling from excess letdown and RCDT heat exchanger 11. RCDT vent 12. Demineralized water to containment 13. Service water to RCP motor air coolers 14. Containment purge and mini purge 15. Containment sump pump discharge 16. Containment sump pump sample recirculation 17. Charging pump / residual heat removal relief 18. Valve discharge to PRT 19. Containment leak rate test 20. Service air 9

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.5.A Dedicated Penetrations for External Recombiner or Post-Accident External Purge System The Farley Plant has redundant electric hydrogen recombiners permanently located in the containment for use in removing hydrogen gas in the containment atmosphere under post-accident conditions. These recombiners meet the engi-neered safety feature requirements; the controls and instrumentation for each are located on separate panels in the main control room. In addition, the post-accident venting system is provided as a backup systen. for the redundant hydrogen recombiners. It consists of a supply line through which pressurizing air may be admitted to the containment and an exhaust line through which hydrogen-bearing gases may be vented from the containment. The gases are filtered to limit radioactive discharges. 10

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.5.C Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant Farley Nuclear Plant has redundant electric hydrogen recombiners located inside containment for use in removing hydrogen gas from the containment atmosphere during post-accident conditions. These recombiners meet all engi-neered safety feature requirements and the controls and instrumentation for each are located on separate panels in the main control room. The emergency procedure for loss of coolant accident contains detailed instructions for operating the recombiners. Since this system is located inside containment and does not require mechanical hookup after an accident, personnel exposure during use is not a consideration at Farley. Alabama Power has reviewed and upgraded the emergency procedure for operation of electric hydrogen recom-biners located inside containment. Training on this procedure revision has been incorporated into the Unit 2 operator training program. 11

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.6.A Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs Alabama Power Company has instituted a program to maintain leakage rates of systems outside containment to as low as practical which consists of the following: A. Systems included in the program: 1. High head safety injection system (recirculation portion only). 2. Low head safety injection system (recirculation portion only). 3. Residual heat removal system. 4. Reactor coolant system letdown and makeup system. 5. Reactor coolant sampling system. 6. Containment spray system (recirculation portion only). 7. Radioactive waste gas system. B. Systems excluded from the program: 1. Radioactive liquid waste system (excluded by NRC in regional meeting). 2. Radioactive waste gas syster. (Portions of system not contaminated by volume control tank off gas processing will be excluded. Off gas processing would be the means of handling highly radioactive gases resulting from various accidents.) C. Procedures for determining (measuring) leakaoe: 1. High head safety injection system - integrated leak rate test. 2. Low head safety injection system - integrated leak rate test. 3. Residual heat removal system - integrated leak rate test. 4. Reactor coolant system letdown and makeup system - integrated leak rate test. 5. Reactor coolant sampling system - integrated leak rate test. 6. Containment spray system - integrated leak rate test. 7. Radioactive waste gas system - snoop test. D. Actual leakage rates will be determined during plant startup testing prior to initial criticality and reported to the NRC. As part of the preventive maintenance program on the scoped systems, leak rate measurements will be performed periodically at intervals not to exceed each refueling outage. 12

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.6.B Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations Previous shielding design studies did not take into account the source terms required by the NRC. A shielding review has been initiated to identify areas which need improved shielding. This shielding review will use NRC source terms. The review will identify areas of requiring access for operation of essential safe shutdown equipment. The radiation sources in these areas will be identified by a field walkdown or design review. The design review will be completed prior to receipt of an operating license. Aspects of all major acci-dents will be covered in the design review, and shielding additions developed in the design review will be implemented by January 1, 1981. 13

Farley huclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.7.A Automatic Initiation of the Auxiliary Feedwater System The requirements of NUREG-0578 are met by the Farley Nuclear Plant as shown by the following: Position 1: The design shall provide for the automatic initiation of the auxiliary feedwater system.

Response

With the control switches in the auto position, the motor-driven auxiliary feedwater pumps will automatically start on any of the following signals: 1. Tripping of both steam generator feed pumps. 2. Low-low water level signals from two-out-of-three level transmitters on any one steam generator. 3. Any of the conditions as defined in FSAR Section 7.3 that cause a safety injection signal. 4. Loss of offsite power. Operation of the turbine-driven auxiliary feedwater pump is initiated by the opening of the steam supply valves tu the turbine drive. Steam from the main steam header is auto-matically admitted to the turbine drive on either of the follow-ing signals: 1. Loss of power signal (2/3 reactor coolant pump bus under-voltage), or 2. Low-low water level signals from two sut-of-three of the level transmitters of any two-out-of-three steam generators. Details of the emergency operation of the auxiliary feedwater systems are contained in FSAR Sections 6.5.2.2.5 and 6.5.2.3.3. Instrumentation and controls for the system are shown on draw-ings D-205007, D-205033, D-207186, D-207188, 0-207189, 0-207590, 0-207591, and D-207857. Position 2: The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxil-iary feedwater system function.

Response

The auxiliary feedwater system is designed to maet the single failure criteria so that no single failure will prevent the supply of sufficient feedwater to at least two of the three steam generators. A detailed design evaluation is contained in FSAR Section 6.5.3. A failure analysis of the auxiliary feed-water system is provided in Table 6.5-2 of the FSAR. Position 3: Testability of the initiating signals and ci:cuits shall be a feature of the design. 14

Farley Nuclear Plant Unit 2 Docket No. 50-364 Section 2.1.7.A

Response

In order to ensure the operability of the auxiliary feedwater system, periodic testing of the system is performed in accord-ance with the Farley Nuclear Plant Standard Technical Specifi-cations, Section 3/4.7.1.2. These technical specifications list the limiting conditions for operation and the surveillance requirements for the auxiliary feedwater system. These surveil-lance requirements ensure that both the motor-driven pumps and turbine-driven pump are operable and develop at least 93 percent of design discharge pressure; that each valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and that the motor-driven and turbine-driven pumps start automatically upon receipt of a test signal which simulates emergency operation of the system. Position 4: The initiating signals and circuits shall be powered from the emergency buses.

Response

The initiating signals and circuits for the auxiliary feedwater system are powered from the emergency buses, D-207001, D-207005, and D-207006. Position 5: Manual capability to initiate the auxiliary feedwater system from the control roem shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system functions.

Response

The auxiliary feedwater system can be operated locally from the hot shutdown panel or remotely from the control room. The operation of the system is described in FSAR Section 6.5.2.3. The system is designed to meet the single failure criteria and a failure analysis of the system is contained in FSAR Table 6.5-2. Position 6: The ac motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simulta-neous and/or sequential) of the loads onto the emergency buses.

Response

The ac motor-driven pumps and valves in the auxiliary feedwater system are included in the automatic actuation of the loads onto the emergency buses as indicated on the emergency load sequence drawings 0-207001, D-207005, 0-207006, D-207645, 0-207646, D-207649, and D-207650. Position 7: The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.

Response

The auxiliary feedwater system can be operated locally from the hot shutdown panel or remotely from the control room as detailed in FSAR Section 6.5.2.2.5. Instrumentation and controls for the system are shown on P & ID drawing D-205007 and the initiating logic on Figure 7.2-14 of the FSAR. 15

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.7.B Auxiliary Feedwater Flow Indication to Steam Generators for PWRs Auxiliary feedwater injection lines to each steam generator are provided with flow indication. This flow indication is on the main control board and is powered from the plant emergency power. These flow instrument loops are testable. Redundancy requirements are met by qualified steam generator level instrumentation (safety grade). A description of the presently installed equipment is contained below. The auxiliary feedwater line flow indicators will be seismically and environ-mentally qualified by January 1, 1981. This will meet all safety grade re-quirements. Position 1: Auxiliary feedwater flow indication to each steam generator shall satisfy the single failure criterion.

Response

Local and control room indication of auxiliary feedwater flow to ecch of the steam generators is provided by flow orifices in the auxiliary feedwater supply lines, located just upstream of the auxiliary feedwater stop check valves. The auxiliary feedwater flow indication is backed up by redundant safety grade steam generator level channels. Instrumentation and controls for the system are shown on P & ID drawing 0-205007. Positiu, 2: Testability of the auxiliary feedwater flow indication channels shall be a feature of the design.

Response

Testing of the auxiliary feedwater flow indication is performed on 18-month intervals by injection of a test signal at the primary sensor. The instruments are calibrated if the output signals do not meet the required accuracy for the instrument. Position 3: Auxiliary feedwater flow instrument channels shall be powered from the vital instrument buses.

Response

The auxiliary feedwater flow instrumentation channels receive their power from the vital instrument bus. The Bechtel Corpora-tion drawing A-207076 gives a block diagram of the steam genera-tor feedwater flow indication loop and the power supply is indica-ted on drawing D-207024 and D-207025. Position a: Each auxiliary feedwater channel should provide an indication of feed flow with an accuracy on the order of 10 percent.

Response

The present control grade system has transmitters with a 0.5 percent full scale accuracy. The Westinghouse power supply has a gain accuracy of !0.1 percent of full scale and the control board indicator has an accuracy of 1.5 percent of full scale. The additive accuracy of the flow loop is 2.1 percent of full scale which is well within the 10 percent range. 16

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.8.A Post-Accident Sampling Capability Methods of obtaining highly radioactive samples of the reactor coolant (i.e., pressurized and unpressurized) and containment atomsphere have been estab-lished for use when Unit 2 becomes operational. The reactor coolant sample will be obtained by using the existino sample system with modifications to allow for remote operation of the samp k valves and transportation of the shielded sample to a shielded lab area for chemical analysis and sample dilu-tion. The containment atmosphere sample will be obtained by modifying the existing monitoring system to allow particulate and radiciodine samples to be taken using small gas volumes while minimizing personnel exposure. The par-ticulate filter and silver 7eolite radiciodine sample will be transported in a shielded container to a remote temporary area for Ge(Li) gamma ray spectro-scopic analysis. Procedure modifications for handling and analysis of sam-ples, plant modifications, and a design review of the sample system are being completed. Additional information on these modifications is contained in the system operation description which follows. The procedure and plant system modifications described above will be completed prior to receipt of an opera-ting license. The modified operational sampling system will also meet the January 1,1981 requirements 17

Farley Nuclear Plant Unit 2 Docket No. 50-364 ATTACHMENT (SECTION 2.1.8.A) Description of Post-Accident Sampling Systems Reactor Coolant (Pressurized and Unpressurized) The Unit 2 reactor coolant sampling system will be modified by addition of a post-accident sampling panel and shielded sample pass-through (Figure 2). A schematic of the post-accident sampling panel is shown in Figure 3. The above system has the capability for remotely taking a pressurized sample (RCS) and an uipressurized sample (RHR/ containment sump). Based upon the results of the shielding design study with subsequent modifications (2.1.6.B) and time studies for drawing the samples, the estimated whole body or extremities radiation doses to any individual will not exceed 3 and 18 3/4 rems, respectively. Vent Stack Effluent The Unit 2 vent stack air particulate and noble gas monitor (RE 21 & 22) will be modified by installation of a post-accidant particulate and iodine sampler, as well as addition of a septum for drawing a gas sample (Figure 4). Sub-sequent to an accident the normally used radiation monitor (RE 21 & 22) will be valved out and the effluent flow directed through the post-accident sampler for a specified time period by operation of the remote control panel. Based upon the results of the shielding design study with subsequent modificaticns (2.1.6.B) and sampling time studies, the estimated whole body or extremities radiation doses to any individual will not exceed 3 and 18 3/4 rems, respec-tively. Containment Air The Unit 2 containment air particulate and noble gas monitor (RE 11 & 12) will be modified by installation of a post-accident particulate and iodine sampler, as well as addition of a septum for drawing a gas sample (Figure 5). Sub-sequent to an accident the normally used radiation monitor (RE 11 & 12) will be valved out and the containment air flow directed through the post-accident sampler for a specified time period by operation of the remote control panel. Based upon the results of the shielding design study with subsequent modifi-cations (2.1.6.8) and sampling time studies, the estimated whole body or evtremities radiation doses to any individual will not exceed 3 anf 18 3/4 rems, respectively. 18

SIDE VIEW REMOTE CONTROL PANEL T ] POST ACCIDENT SAMPLING PANEL ()

• REMOVABLE SHIELDED PLUG 7lF

- LEAD SAMPLE PlG IS MOVED FORWARD ON TRACK O b TRACK ON CART MATES UP TO THACK PERMANENTLY MOUNTED L IN CORE DRILLED PASSAGE 7 w* PIG COMES TO REST HERE,WHERE LIQUID SAMPLE IS DISPENSED TO SAMPLE VIAL O O EAST WALL OF SAMPLE ROOM Figure 2 Reactor Coolant Sampling System 1840-0

p________q l T V -2 I TO VACUUM 1 I ) L "V _J . _gg _p r- - - - - - - - -, SV8 Tv. g GAGE ARGON NORMALLY SUPPLY N CC CLOSED SURGE DOMB SV7 SEPTUM LINE FOR GAS SAMPLE O GAS EXPANSION BOMB SAMPLE HETURN LINE ARROWS INDICATE THE DEENERGlZED FLOW PATH 6 to RHR/RCS o HOT LEG SV1 VCT GAS SV4 SPACE SV5 XDV1 ,J SV3 [ STEAM LIQUID SAMPLE LINE DRA T TO LE AD SHIELD Sv2 k P2R LIQUID Figure 3 Reactor Coolant Post-Accident Sampling Panel 1840 4

r/ CONTROL PANEL ~ MOUNTING BO ARD FOR SAMPLE SYSTEM SAMPLE SAMPLE CONTAINER SOM FILTE R a SV1 U WALL I~~~~~~~~J SAMPLE F ROM I g VENT STACK

jG' -

I I R E 21 & 22 [ RETURN TO ~ l ( s g l g VENT STACK X f I I Y I I I I I I ( I k-----.__) o s / PUMP -A Figure 4 Vent Stack Effluent Sampling System 21 1840-0

SAMPLE CONTAINER SHtELD WALL FILTER ASS'Y (TO BE INSTALLED) WITH QUICK Syg DISCONNECT r% SV1 WALL EXISTING RE 11 & 12 EXISTING PUMP OUTLET ) { rh { OUTLET SAMPLE LINE CONTAINMENT 1 INLET SAMPLE LINE Figure 5 Containment Air Sampling System 22 1840-0

Farley Nuclear Plant Unit 2 Docket No. 50-364 SEC110N 2.1.8.8 Interim Procedures for Quantifying High Level Accidental Radioactivity Releases To measure the noble gas radioactive effluent release, Alabama Power Company will mount a Jordan rad gun on the vent stack at the 175-foot level. This monitor will use a shielded isokinetic sampler to obtain a representative sample and lead shielding to reduce background interference; this will provide a continuous readout at th monitor. The monitor will be DC powered with a battery life in excess of 30 days. Monitor readings will be provided to the control room via verbal communications using any one of the three existing communications systems: (1) plant phone system; (2) sound powered phone system; or (3) plant public address. The monitor has a range of 0.01 mR/hr to 10,000 R/hr over an energy range of 80 kev t23).2 MeV. Calibration will be done at installation and annually using a Cs calibration source. Predeter-mined calculational methods will be used to convert the radiation level read-ing to radioactive effluent release rate. The measuren. cot of radiciodine and particulate effluents will be accomplished by a modification to the normal vent stack monitor RE 21 & 22 which allows the collection of small gas samples on particulate filters and silver zeolite cartridges, which are analyzed using a Ge(Li) gamma ray spectroscopy system. Procedures for operation of the system have been developed to provide calcu-lational methods to determine release rates. These modifications will be com-pleted prior to receipt of an operating license. (NOTE: If the requirements of Section 2.1.8.B1 are met prior to receipt of the Unit 2 operating license, the above interim procedures will not be used.) 23

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.8.B1 High Range Effluent Monitor To meet the January 1,1981 requirements for a high range noble gas effluent monitor,AlabamaPowerCompanyhasorderedanEberlineSPING-4samp}er. This sagpler will monitor the vent stack effluent and has a range of 10 pCi/cc to 10 pCi/cc by using multiple ranges for noble gases. The monitor readout will be located in the control room area and will be powered from a vital instru-ment bus. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage. Procedures will be developed for use, calibration of the system, and dissemination of release rate information. 24

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.8.B2 High Range Effluent Radioiodine and Particulate Sampling and Analysis To meet the January 1, 1981 requirements for a high range effluent radioiodine and particulate sampling system, Alabama Power Company has ordered an Eberline SPING-4 sampler. This sampler provides the capability to monitor ef fluent radioactivity in the form of noble gases, radiciodines, and particulates by use of individual channels for each type of radioactivity. The sampler will monitor the vent stack with the monitor readout located in the control room area. The sampler will be powered from a vital instrument bus. The particu-late channel uses a filter paper in the air stream which is counted by a beta scintillation detector with an alpha detector for subtraction of the radon 5 The range of ge channel is 2.6 x 10 thoron daughter activity contribution. y counts per minute per microcurie on the paper for Cs The radiciodine channel monitors a silver xeolite cartridgin the air stream with a sensitiv-ity of approximately 80k CPM per pCi of I Goth the particulate channel and the radiciodine channel use external sources for calibration and can be compensated for background radiation. Calibration will be performed upon installation and at intervals not exceeding each refueling outage. Procedures will be developed for use of the system, calibration of the monitor, and dissemination of release rate information. 25

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.8.B3 High Containment Radiation A thorough evaluation has been made of all the equipment available at this time and projected to be available during the next year. This evaluation was made by a point-by point comparison of all published material and after con-siderable discussions with each vendor. None of the equipment designs are finalized yet. Also, none of the vendors have completed the environmental qualification tests or performance tests. It is quite likely that the design will have to be modified, perhaps considerably, in order to meet both the NRC performance requirements and environmental qua ifications simultaneously. Some of the equipment has been shown not to meet the performance requirements, while others fail the environmental qualifications. Some information in vendor brochures has been shown by tests to date to be incorrect. Vendors have had to retract certain published reports on performance and revise their figures drastically in the nonconservative direction. Of particular note, none of the present monitor designs can meet the temper-ature requirements for LOCA conditions, and also, temperature ratings in brochures have been reduced. As a result, there is no available equipment that is known to be able to satisfy both sets of requirements (performance and environmental conditions) simultaneously. Therefore, a commitment at this time to buy specific equipment from a particular vendor with the presently proposed designs with known and possible flaws or shortcomings is definitely premature. It is not likely that simple design co.nges will make any of the designs acceptable. Therefore delivery dates and perhaps even design concepts may have to be changed drastically. The present technical specifications require calibration of monitors during each refueling outage. Calibration of these high range containment radiation level monitors will require removal of the detectors from the containment to a shielded calibrator in a calibration room. At this time there are no cali-brators available to calibrate the upper decades of these monitors. For safety purposes, it would be better to have the monitors calibrated at some centrally located facility offsite. Alabama Power commits to installing a containment radiation monitor with the range specified in NUREG-0573 upon availability of a production type monitor. 26

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.8.C In-Plant Iodine Instrumentation Alabama Power Company has a portable monitoring system available which uses an iodine silver xeolite sampler and single channel analyzer. Emergency pro-cedures have been revised to address the use of this portable monitor. Appro-priate shift personnel have been trained on the use of this analyzer. Alabama Power Company presently has the capability of purging these samples of entrapped noble gases by the use of nitrogen gas and analysis by Ge(Li) gamma ray spectroscopy in a low background counting facility. 27

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.1.9 Transient and Accident Analysis Analyses of small break loss of coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and analysis of tran-sient and accident scenarios, including operator actions not previously ana-lyzed, are being performed on a generic basis by the Westinghouse Owners' Group, of which Alabama Power Company is a member. The small break analyses have been completed and were reported in WCAP-9600, which was submitted to the Bulletins and Orders (B & 0) Task Force by the Owners' Group on Jur.e 29, 1979. Incorporated in that report were guidelines that were dr. eloped as a result of small break analyses. These guidelines have been reviewed and approved by the B & 0 Task Force and have been presented to the Owners' Group utility repre-sentatives in a seminar held on October 16-19, 1979. Following this seminar, Alabama Power has revised emergency procedures and has trained personnel on these procedures. Revised procedures and training are in place in accordance with the requirement in Enclosure 6 to Mr. Vassallo's letter of September 27, 1979 and included in the Unit 2 operator training program. The work required to address the other two areas--inadequate core cooling and other transient and accident scenarios--has been performed in conjunction with schedules and requirements established by the Bulletins and Orders Task Force. Analysis related to the definition of inadequate core cooling and guidelines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation and for restoring core cooling following a small break LOCA were submitted on October 31, 1979. This analysis is a less detailed analysis than was originally proposed and will be followed up with a more extensive and detailed analysis which will be available during the first quarter of 1980. The guidelines and training are included in the Unit 2 operator training program required by the B & 0 Task Force. With respect to other transient accidents contained in Chapter 15 of the J. M. Farley FSAR, the Westinghouse Owners' Group has performed an evaluation of the actions which occur during an event by constructing sequence of event trees f or each of the non-LOCA and LOCA transients. From these event trees a list of decision points for operator action has been prepared, along with a list of information available to the operator at each decision point. Follow-ing this, criteria have been set for credible misoperation, and time available for operation decisions have been qualitatively assessed. The information developed will then be used to test Abnormal and Emergency Operating Procedures age?nst the event sequences and determine if inadequacies exist. The results of this study will be provided to the Bulletins and Orders Task Force on March 31, 1980 as required. Alabama Power Company will revise procedures 'and complete retraining p:ior to receipt of an operating license. The Owners' Group has also provided test predictions analysis of the LOFT L3-1 nuclear small break experiment. This analysis was provided on December 15, 1979 in accordance with the schedule established mutually with the Bulletins and Orders Task Force. 28

Farley Nuclear Plant Unit 2 Docket No. 50-364 Containment Pressure Indication (ACRS) The present containment pressure indication provides continuous redundant indication in the main control room and has an indication range of -5 psig to 60 psig. Additional monitoring capability with control room indication having a range of 0 to 210 psig will be installed by January 1, 1981. This addi-tional monitoring equipment will be safety grade and meet the design pro-visions of Regulator Guide 1.97, Revision 1. 29

Farley Nuclear Plant Unit 2 Docket No. 50-364 Containment Water Level Monitor (ACRS) The Farley Nuclear Plant present design has two wide range containment water level detectors. These detectors provide indications in the main control room that meet the wide range requirements as specified in Mr. Vassallo's September 27, 1979 letter. These level transmitters and associated readout are safety grade and measure volumes in excess of 600,000 gallons. The Farley Nuclear Plant will provide a narrow range containment level indica-tion meeting the ranges specified in Mr. Vassallo's September 27, 1979 letter by January 1,1981. 30

Farley Nuclear Plant Unit 2 Docket No. 50-364 Containment Hydrogen Indication (ACR5) Two independent, redundant systems for containment hydrogen monitoring are provided in the present design with a range of 0 to 10 percent hydrogen concen-tration. The design of these systems follows, as applicable, the requirements for ~'ety-related protective systems and meets the requirements of IEEE 279-1971. The output signal of the analyzers are indicated at the analyzer panel loca-tion and are recorded and alarmed in the main control room. Each system i5 supplied electrical power from an independent and redundant Class IE pover supply. The system meets the single failure criteria and remains operable under pos-tulated accidents. Any single failure in one hydrogen monitoring system does not affect its redundant and independent counterpart. The system described above currently meets the January 1,1981 requirements. 31

Farley Nuclear Plant Unit 2 Docket No. 50-364 Reactor Coolant System Venting Alabama Power Company will install a reactor vessel head vent system by Janu-ary 1, 1981. The following is a description of the system proposed:

System Description

The reactor vessel head vent system is designed to remove gases from the reactor ccolant system via remote manual operations from the main control The reactor vessel head vent system will discharge into a well-venti-room. lated area of the containment in order to ensure optimum dilution of com-bustible gases. Inside the containment hydrogen can be recombined by means of the post-accident hydrogen recombiners. The discharge point will be designed for adequate drainage of reactor coolant in the case of inadvertent dis-charges. The reactor vessel head vent system flow diagram is shown in Figure 6. The system arrangement provides for venting the reactor vessel head by using only safety grade equipment. The system mainly consists of 1-inch piping with four safety Class 2 " fail closed" isolation valves. The system utilizes all nor-mally closed valves. The isolation valves 1 and 2 are powered by train A and the isolation valves 3 and 4 are powered by train 8. The system is designed such that any single active failure will not prevent vessel gas venting nor prevent venting isolation. The system also provides the necessary manual venting functions during vessel filling operations. The system connects to the reactor vessel head at the existing vent pipe with redundant flow paths through orifices. The orifice restricts the flow rate f rom a pipe break downstream of the orifice to within ti:e makeup capacity of one centrifugal charging pump. All piping and equipment between the orifice and the discharge point is of Safety Class 2 as defined by ANSI N18.2A. The reactor vessel head vent system isolation valves will be supported from the seismic support platform. The system can be disconnected downstream of the second isolation valve to accommodate refueling. In this manner the necessary flanged connections will be outside of the reactor coolant pressure boundary. All piping and valves upstream of the flanges will remain integral with the reactor vessel head at all times. 32

TO CONTAINMENT o Em Mc 1 3 S S EXISTING VENT LINE ORIFICES REACTOR VESSEL l HEAD Figure 6 Flow Diagram of the Reactor Vessel Head Vent System 33 18404

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.2.1.A Shift Supervisor Responsibilities 1. Corporate management has issued and will periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions on his shif t and that clearly establishes his command duties. 2. Plant procedures have been revised to assure that the duties, respon-sibil.ities, and authority of the shif t supervisor and control room oper-ators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shif t supervisor in the control room relative to other plant management personnel. 3. Tra,ining programs for shift supervisors emphasize and reinforce the re-sponsibility for safe operation and the management function the shif t supervisor is to provide for assuring safety. 4. Corporate management has reviewed the administrative duties of the shift supervisor. Administrative functions that detract from safe operation have been delegated to other personnel. 34

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.2.1.B Shift Technical Advisor An individual on shift has been designated to serve in a dedicated capacity as shift technical advisor during emergency conditions. This individual reports to the shift supervisor during emergency conditions and serves in a technical advisory capacity. Personnel designated for this function are not part of the minimum shift complement, in accordance with technical specifications. Desig-nated individuals have other duties during normal plant conditions. Alabama Power has placed shift technical advisors on shift. By January 1, 1981, the shif t technical advisor will have received additional training and periodic retraining, commensurate with NUREG-0578 requirements, in the basic physical concepts of mathematics, chemistry, metallurgy, atomic physics, reactor physics, heat transfer, fluid mechanics, thermodynamics, and thermo-hydraulics. In addition, training will be provided in the response and anal-ysis of the plant for various transients and accidents and in basic plant design and layout including the capabilities of instrumentation and controls in the control room. The operating experience assessment function will be performed by the plant's system performance group which is (omposed of supervisory, engineering, and technical personnel. This group is not functionally a part of the plant operations group. The system's performance group is a multidisciplined group which has overview of all plant systems including mechanical, electrical, and instrumentation and control. This group is Jedicated to the operating experi-ence assessment function which includes but is not limited to the following: 1. Engineering evaluation of the operating history of the plant (equip-ment failures, design problems, operations errors, etc.) and Licensee Event Reports from other plants of similar design, with suitable dissemination of the results of such evaluations to other members of toe plant staff. 2. Engineering evaluation of the adequacy of the policy for maintenance, testing, equipment procurement, etc. 3. Engineering evaluation of continuing adequacy of plant operations quality assurance. 4. Engi:,ering evaluation of adequacy of plant emergency and operating procedures. The operati"g experience assessment function will be implemented upon receipt of an operating license. Shift technical advisors and personnel performing the operating experience assessment function will communicate and appropriately advise each other of pertinent safety concerns. 35

Farley Nuclear Plant Unit 2 Docket No. 50-364 LECTION 2.2.1.C Shift and Relief Turnover Procedures Plant procedures for shif t and relief turnover have been reviewed and revised to assure that: 1. A checklist has been provided for the shift relief of control room operators and the shift supervisor. The checklist includes the following items: a. Assurance that critical plant parameters are within allowable limits. b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of oper-ational transients and accidents. c. Review of systems and components that are in a degraded mode of operation permitted by the technical specifications. 2. " heck'ists or logs have been provided for shift relief of assistant plant operators and equipment operators. Equipment that could de-grade a system critical to the prevention and mitigation of oper-ational transients and accidents or inititate an operational tran-sient has been listed on the appropriate operators' checklist. 3. A system has been established to evaluate the effectiveness of the turnover procedure. 36

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.2.2.A Control Room Access Plant procedures have been reviewed and revised to assure the following: 1. The authority and responsibility of the persons in charge of limiting access to the "at the controls area" has been established. 2. A clear line of authority and responsibility in the control room in normal and emergency conditions has been established. The line of succession for persons in charge of the controi room has been estab-lished. Lines of communication and authority for plant management personnel not in direct command of operation are defined. 37

Far ley Nuclear Plant Unit 2 Docket No. 50-364 SEC' W 2.2 2.B Onsite Tecar a' Suoport Center An interim technical support center (TSC) has been established to meet require-ments for receipt of an operating license (Figure 7). The plant emergency plan and appropriate implementing procedures have been revised to incorporate the role and location of the interim cechnical support center as part of the emergency preparedness improvements. The following is a description of the proposed long-term onsite technical support center to meet the January 1, 1981 requirements. This design covers the anticipated manning requirements, document storage, monitoring systems, communication, habitability, and structural arrangements. The center has been designed to be used for post-accident and normal plant conditions. Conceptual Design 1. Location - The TSC is located in the Auxiliary Building, elevation 155', immediately north of the Unit 2 control room area. Personnel will normally enter and exit via the primary security access point and Unit 1/ Unit 2 control room area. Passage through the control room will not interfere with the operators who are manning the control boards. An emergency or alternate exit is from the north side of the TSC and throu*gh an exit door in the northwest wall of the Unit 2 Auxiliary Building. 2. Size and Room Layout - The TSC is designed to accommodate 25 people for the performance of post-accident monitoring, evaluation of plant status, coordination of damage assessment and emergency actions, interface with the NRC and emergency operations center, and onsite/offsite communications. Figure 8 shows a proposed room layout. An overall space of 22 feet x 65 feet, with a 9-foot ceiling height, has been provided. Room layout is as follows: a. Monitoring Area - Includes CRT, trend line printer and alarm type-writer for each unit, and provision for communication with the control room, emergency operations center, NRC, and the State of Alabama, b. Planning and Coordination Area - Includes desks, reference tables, files for plant procedures and manuals, and monitors for TV cameras in the control room. Tables and additional folding chairs and movable partitions will be provided so that the area can be quickly set up for usage. Phones will be provided for full communication capability. A chalk board and projection screen will also be located here. Figures 9 and 10 describe the TV monitoring system. c. Document Room - Includes files for microfilm, drawings, data sheets, indexes, and microfilm reader and printer. d. Storage Area - Includes cabinets for paper, office supplies, etc. 38

Farley Nuclear Plant Unit 2 Docket No. 50-364 Section 2.2.2.B 3. Construction - Structural and e 'cetural design criteria of the TSC will be similar to that used for w ecfice areas in the control room, except that it will not meet saivic catagory 1 requirements. 4. Activation - Activation is in accordance with the Emergency Plan. Equip-ment will be capable of displavina vital plant parameters from the time an accident begins; TV cameras and u tars will be turned on as needed. 5. Monitoring and Display Equipme t - 9.e philosophy of data monitoring is that a person would be located at the main control room computer console and communicate with the TSC via sound powered phones. TSC personnel would request display of the desired data in the TSC. One of the two CRTs normally located in the main control room (ene per unit) would be discon-nected, rolled into the TSC, and booked up upon activation of the TSC. A two pen recorder located in the TSC would have the capability to trend two parameters. The line printers (one per unit) now 1::ated in the computer room (Auxiliary Building, elevation 121') would be moved to the TSC. Parameters could be trended on the line printers providing a hard copy of data for review. 6. Communications - Communications will be provided to contact the NRC, State of Alabama, and APC General Office. Sound powered phones will be provided between the TSC and main control room. An intercom system will be avail-able between the T5C, the three operations support centers, and the emer-gency operations center. Commercial and APC telecommunications lines will be available for redundant offsite communications. An APC security radio will also be located in the TSC. 7. Habitability - The TSC will not be exposed to any areas in the Auxiliary Building that could contain highly radioactive sources post-accident. The ventilation system will include a deep-bed charcoal filter to remove airborne contamination, and it will have the capability of pressurizing the TSC area and recirculating the room air through the charcoal filter. A permanent radiation monitor will be provided to continuously indicate radiation dose rates and airborne activity. A radiation alarm in the makeup air supply duct will automatically initiate room pressurization and recirculation. The HVAC system is described in detail under system de-scription. 8. Fire Protection - The TSC is enclosed by 3-hour fire-rated walls with Class A airtight fire doors. The present area, room 2452, is provided with fire and smoke detectors and a wet pipe sprinkler system. A water hose cabinet and CO h se reel are presently located within the TSC area. 2 One CC and one dry chemical fire extinguisher are provided inside of the 3 TSC. i.mergency breathing apparatus and spare air bottles are now provided for the control room personnel. 9. Electrical Power - Normal and emergency provisions are as follows: 39

Farley Naclear Plant Unit 2 Docket No. 50-364 Section 2.2.2.8 a. Closed Circuit TV and Computer Hardware - Regulated voltage distri-bution panels 1A and 2A, channel 1, or panels 18 and 28, channel 2. b. HVAC - 600-V MCCs-1F and 1G. These are shared MCCs and can be powered from the diesel generators upon LOSP. c. Wall Receptacles - 120/208-V distribution panels 1R, 15, IEE, or 1FF. These are shared panels and can be powered from the diesel generators upon LOSP. d. Lighting - 26 fluorescent fixtures, four 40-W lamps each; 277-Vac from MCC-1F or 1G (Unit 2) and MCC-2CC or 200 (Unit 2); 25-W emergency lighting; 8-hour battery packs, two 2-head, two 1-head; tied to MCC-200 (Unit 2), train B. e. Wall Receptacles - Fed from MCC-2E (Unit 2), " normal" train; 125-Vac, 15-A. HVAC System Description This system is composed of one (6.4-ton) air conditioning unit with air-cooled condenser, one emergency filtration unit equipped with a separate fan, two 6-inch butterfly valves for positive isolation from contaminated outside air, and associated ductwork and controls. Refer to Figure 11. A positive pressure is maintained in this area by using outside air which is vented directly to the outside from the occupied area. Upon receipt of a high radiation signal from a sensor in the outside air duct, butterfly valves and dampers change position and the outside air plus an equal quantity of recir-culated air is directed through the emergency filter, thus maintaining the pressure within the conditioned space and establishing a cleanup mode of operation simultaneously. The temperature is maintained by a roem thermostat which cycles either the compressor or the electric heating coil within the air conditioning unit upon demand. The filtration unit contains a 50 percent prefilter, a 99.97 percent HEPA filter, plus a carbon filter. 40

M N CONTROL ROOM SUPPLY r. KITCHEN SUPPLY m q l l l INSTRUMENT R ACKS KITCHEN i TOILET / MAIN p 4,0 i TECHNICAL 7 SUPPO RT CENTER g, (INTERIM) si y g@ O I F OR EMAN'S Il 2 N OF FICE n 11 Il 5 A il 3 [ U 3 t g / 2 OFFICE 0 BOARD y CCZZ CONTROLS AREA I = I I l l n INSTRUMENT RACKS I I i I i 1 s 3 OPE R ATION AL SUPPORT g CENTER x 3 L Figure 7 Control Room Interim Technical Support Center 41 1840-0

a. s i I g 8 i // T / NE M / A / MO E / UO R / CR A / O G / D N / I / R / O L. / IT O / N C / E O / G A A M T O S E XI H R A E TS / r // /f///I' / e t / V n D /l e C E / + T / A / n t O r T R / o A H E GE / R 8 p R 8 R / H 4 p A TI / ) 2 u F M R. / A L O S E I R / E L G A / O A ( H / R A l R H U2 / A W / a 0 / L c O M )L / N T O A I / T C / I H i O R n 5 S L N "2 A / 6 T G T h W / N A 1 T c / N O E / I R / C e C . // D IAED L / T L / R / T O / O (JA / a 8 t C / O hR / C T V CE / e S R / 4 r 1 ' r I G 7t l! u X N g E I / i N / N / F Y / / A R / L /'/ '-/h T = / P N / E / N Y I A TR M SO IXOh ED /. O ////// // / / // A M

e M N CONTROL ROOM SUPPLY r KITCH E N SUPPLY g I I I I I I = q INSTRUMENT R ACKS K ITCH E N g i TOILET / MAIN p TECHNICAL Of SUPPORT l CENTERL g, CCTV 3 (INTERIM) ll 3 ll 2 ll FO R EMAN'S OFFICE te N 11 b ( I!_ 5 o ii it ,_s* 6"_ 9 a L l 0 CCTV 2 3 CCTV 1 O F FIC E 4 4 SYMBOLS ( BOARD y M CCTV CAMER A' WITH PAN & TILT l I I I I i ~ INSTRUMENT RACKS II I i 1 l y a 8 s 5 Figure 9 Control Room CCTV Monitoring Systen 43 1840-0

F i CAMERA 1 l l 1 1 I i 14" l MONITOH I 4* 1 I I I sissisissisis sisilississisis I /////////////////s4/sss///// l / l g a l I I i e / 1 I I I CAMERA 2 e i g u / / \\ i i 3 4.. I l PAN & TILT l MONtTOR VIDEO o I AND 2OOM

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l J ~ sississississisis ris LENS '/ / / / / / / / / / / / / / / / / / 4 rr / l 1 I a s l g s / zu l l L. i y l / I I e CAMERA 3 r l VIDEO I l 14" 14" + TAPE o l MONITOR MONITOR I .t ECO R D E R l

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l I sisississisisisisifsisi/is// !s s s s s s / si f / // / / ///j's / / / // / / / l l l l 1 l CCTV MONITOR CONTROL CONSOLE I I I L__.__._________________._____________________J LEGEND: VIDEO SIG N AL s 'i//s; CONTROL / SIGNAL Figure 10 Technical Support Center Control Room CCTV fionitoring System 1840-0

N2V47 K012-N ~~ I l i i CONDENSING l .g a E L.175' 8" O STFY. VALVE p N2V47V001-N VENT TO OUTSIDE rs30 Cm gg h H7 C O.A. \\ j F= C J P 3 TECHNICAL CONTROL } FC SUPPOR T LC C 375 CFM ROOM CENTER 4 A/C UNIT (EL.1551 l N2V47CO28-N l l u /T\\ 2555 CFM k 3 1 2550/2180 CFM -/2 4 I I I FILTER FAN -[*h k N2V47 F027-N n t o /hs' J u FILTER UNIT 750 CFM N2V47001 N NORMAL A +--- EMERGENCY m a a f s l/ N 375 CFM 375 CFM NOTE: DAMPER OPERATORS ARE ELECTRIC Figure 11 HVAC - Technical Support Center 45 1840-0

Farley Nuclear Plant Unit 2 Docket No. 50-364 SECTION 2.2.2.C Onsite Operational Support Center The southeast corner of the control room has been designated as the operations support center for operations and chemistry and health physics personnel (Fig-ure 12). This area is separate from the controls area and will provide rapid response to requests for operations or chemistry and health physics support by the shift supervisor while precluding congestion of the controls area. All other plant support personnel will report to operations support center areas designated in the plant Service Building. Communications exist between these areas and the controls area. The plant Emergency Plan will be revised to re-flect the operational support centers and to establish methods of management and lines of communication as part of the improved emergency preparedness effort prior to receipt of an operating license. 46

M CONTROL ROOM SUPPLY r KITCHEN SUPPLY = l l l l l INSTRUMENT R ACKS KITCHEN g g l TOILET / l MAIN p i TECHNICAL Of SUPPO RT CENTER gg (INTERIM) Il a l I h g FO R EMAN'S OF FICE ti N 18 t !L rr-l! E 3 3 / t = + ez OFFICE ( BOARD p Z _! _ CONTROLS AREA l l l l INSTRUMENT R ACKS i l l l I I 2 3 OPERATIONAL SUPPORT y CENTER x 3 L Figure 12 Control Room Operational Support Center 47 1840-0

,:1 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 DRii'INGS RESPONSE TO SHORT-TERM LESSONS LEARNED: NUREG-0578

Farley Nuclear Plant Unit 2 Docket No. 50-364 LIST OF DRAWINGS TITLE NUMBER PAGE Auxiliary Feedwater D-205007 50 System Main Steam and 0-205033 51 Auxiliary Steam Systems Single Line - Electrical D-207001 52 Auxiliary System (Emergency) 4160 V and 600 V) Single Line - Protection D-207005 53 and Metering 4160 V Switchgear, Bus 2F Single Line - Protection 0-207006 54 and Metering 4160 V Switchgear, Bus 2F Single Line - 120 Vac D-207024 55 Vital and Regulated System A Single Line - 120 Vac D-207025 56 Vital and Regulated System B Block Diagram A-207076 57 Auxiliary Feedwater Pump D-207186 58 4160V No. 2A & 2B Turbine Driven Auxiliary Feedwater Pump Train "C" ($heet 1 of 2) D-207188 59 (Sheet 2 of 2) D-207189 60 Solenoid Valves D-207590 61 Solenoid Valves D-207591 62 Solenoid '!aizes D-207857 63

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