ML20141M937

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Forwards Description of Instrumentation for Detection of Inadequate Core Cooling Sys (TMI Item II.F.2).Marked-up FSAR Pages 1.9-21 & 4.4-37e,indicating Compliance w/NUREG-0737, Clarification of TMI Action Plan Requirements, Also Encl
ML20141M937
Person / Time
Site: Seabrook  NextEra Energy icon.png
Issue date: 02/24/1986
From: Devincentis J
PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To: Noonan V
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM SBN-952, NUDOCS 8603030078
Download: ML20141M937 (64)
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{{#Wiki_filter:SEABROOK STATION Engineering Office Pub 5c SeMee of New HampsNro Februar 24, 1986 New Hampshire Yankee Divlelen T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5

References:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNii Letter, J. DeVincentis to G. W. Knighton, " Compliance with NUREG-0737: Clarification of TMI Action Plan Requirements," dated October 10, 1985

Subject:

NUREG-0737 Task II.F.2, Instrumentation for Detection of Inadequate Core Cooling

Dear Sir:

In Reference (b), we indicated that the Seabrook Station will be equipped with Instrumentation for Detection of Inadequate Core Cooling System. Enclosed herewith please find the description of this system (Attachment 1) and marked-up FSAR Pages 1.9-21 and 4.4-37e which indicate Seabrook's compliance with NUREG-0737, " Clarification of TMI Action Plan Requirements." The marked-up FSAR pages will be incorporated into the FSAR by a future amendment. Should you cr your staff have any questions, please do not hesitate to contact us. We do request that the acceptability of this system be reflet;ted in the next supplement to Seabrook Station's SER. Very truly yours, 1D John DeVincentis, Director Engineering and Licensing Attachments cc: Atomic Safety and Licensing Board Service List 8603030078 860224 O Ml PDR ADOCK 05000443 A PDR 't P.O Box 300. Sootwook. NH O3874. Tolophone (603)474-9521 1 ~

Willico S. Jcrd:n, III Don 21d E. Chick Dicn2 Currca T:wn Hanrg;r Harmon, Weiss & Jordan Town of Exeter 20001 S. Street, N.W. 10 Front Street Suite 430 Exeter, NH 03833 Washington, D.C. 20009 Brentwood Board of Selectmen Robert G. Perlis RED Dalton Road Office of the Executive Legal Director Brentwood, NH 03833 U.S. Nuclear Regulatory Commission Washington, DC 20555 Richard E. Sullivan, Mayor City Hall Robert A. Backus, Esquire Newburyport, MA 01950 116 Lowell Street P.O. Box 516 Calvin A. Canney Manchester, NH 03105 City Manager City Hall Philip Ahrens, Esquire 126 Daniel Street Assistant Attorney General Portsmouth, NH 03801 Augusta, ME 04333 Stephen E. Merrill Mr. John B. Tanzer Attorney General Designated Representative of Dana Bisbee, Esquire the Town of Hampton Assistant Attorney General 5 Morningside Drive Office of the Attorney General Hampton, NH 03842 25 Capitol Street Concord, NH 03301-6397 Roberta C. Pevear Designated Representative of Anne Verge, Chairperson the Town of Hampton Falls Board of Selectmen Drinkwater Road Town Hall Hampton Falls, NH 03844 South Hampton, NH 03827 Mrs. Sandra Gavutis Patrick J. McKeon Designated Representative of Selectmen's Office the Town of Kensington 10 Central Road RFD 1 Rye, NH 03870 East Kingston, NH 03827 Carole F. Kagan, Esquire Jo Ann Shotwell, Esquire Atomic Safety and Licensing Board Panel Assistant Attorney General U.S. Nuclear Regulatory Commission Environmental Protection Bureau Washington, DC 20555 Department of the Attorney General One Ashburton Place,19th Floor Mr. Angi Machiros Boston, MA 02108 Chairman of the Board of Selectmen Town of Newbury Senator Gordon J. Humphrey Newbury, MA 01950 U.S. Senate Washington, DC 20510 Town Manager's Office (ATTN: Tom Burack) Town Hall - Friend Street Amesbury, MA 01913 Diana P. Randall 70 Collins Street Senator Gordon J. Humphrey 'Seabrook, NH 03874 1 Pillsbury Street ) Concord, NH 03301 ) Richard A. Hampe (ATTN: Herb Boynton) i Hampe and McNicholas ) 35 Pleasant Street H. Joseph Flynn Concord, NH 03301 Of fice of General Counsel Federal Emergency Management Agency 500 C Street, SW Washington, DC 20472

J... SBN-952 SB 1 & 2 Amendment 56 FSAR November 1985 s rse sauel,o"l4.u.s. z. Jdihe.,, /abcn for b)eleclocn o et ~ cc.hele descrifltcn 'D ScubnCch %slrumen +,. ggg _ yale &c Ccclin(gajed (; fyyeuy }g jg), .Tw le was-rdauuHed lo fke NRC y,k icket t e Responset, 95g ollowing instrumentation will.be used at Seabrook to detegt ID ' } Reactor Coo nt,laventory Monitor, f Saturation Monitor, and Core Exit'Thermocouples Seabrook Station will be using the Westinghouse design.for-detection of ICC Upon finalization of' plant sper.ific design details, more comprehensive k ~ ~ ~ - ~ ~ ~ ~ ~ g ormation will be provided. Task II.G.1 Emergency Power for Pressurizer Equipment (NUREG-0737) Position: Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17, and 20, of Appendix A to,10 CFR, Part 50, for the event of loss-of-offsite power, the following positions shall be implemented. ^ Power Supply for Pressurizer Relief and Block Valves and Pressuriser Level Indicators (1) Motive and control components of the Power Operated Relief Valves (PORVs) shall be capable of being supplied from either the off-site power source or the emergency pever source when the off-site power is not available. (2) Motive and control components associated with the PORV block valves shall be capable of being supplied from either the off-site power source or the emergency power source when the off-site power is not available. (3) Motive and control power connections to the emergency buses for the PORVs and their associated blocl. valves shall be through devices I that have been qualified in accordance with safety grade requirements. (4) The pressurizer level indication instrument channels shall be powered from the vital instrument buses. The buses shall have the capability of being supplied from either the of f-site power source or the emergency source when off-site power is not available.

Response

l See TSAR Sections 7.5 and 8.3.1. G Fs M Pq c.1.9-21

1 SBN-952 Amendment 55 SB 1 & 2 July 1985 FSAR ( Reg. Guide Discussion Position The procedures for performing channel check, channel functional l C.4.h test, and background noise measurements will be available for l review six months prior to fuel load. Radiation protection procedures have been developed to provide C.4.i guidance and direction to station personnel for minimizing radiation exposure during maintenance, calibration, and j diagnostic work activities. The overall radiation protection program is described in FSAR Chapter 12. Seabrook's non-licensed training program will provide pertinent I j C.4.j general and specific training for plant personnel involved j with system operation, maintenance, and loose part diagnosis prior to station operation. If the presence of a loose part is confirmed and is evaluated C.6 to have safety significance, it will be reported to the NRC 4 (se in accordance with 10CFR50.72. n AHaz%f. 1 y4A.L.S 4.4.7 References Christensen, J. A., Allio, R. J. and Biancheria, A., " Melting Point 1. of Irradiated UO," WCAP-6065, February, 1965. 2 Motley, F. E., Wenzel, A. H. and Cadek, F. F., " Critical Heat Flux 2. Testing of 17 x 17 Fuel Assembly Geometry with 22 Inch Grid Spacing," WCAP-8536 (Proprietary), May, 1975 and WCAP-8537, May, 1975. 1 Hellman, J. M. (Ed.), " Fuel Densification Experimental Results and 3. Model for Reactor Application," WCAP-8218-P-A (Proprietary) March, I l 1975 and WCAP-8219-A, March, 1975. l i Tong, L. S., " Boiling Heat Transfer and Two-Phase Flow," John Wiley 4. & Sons, New York, 1965. ) Tong. L. S., " Boiling Crisis and Critical Heat Flux," AEC Critical 5. i Review Series, TID-25887, 1972. i Tong, L. S., " Critical Heat Fluxes on Rod Bundles," in "Two-Phase 6. Flow and Heat Transfer in Rod Bundles," pp. 31-41, American Society of Mechanical Engineers, New York, 1969.

Chelemer, H., Weisman, J. and Tong, L.

S., "Subchannel Thermal j 7. Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January, 1969. FSN? Ry 4.4-37.

SBN-952 ATTACHMENT TO FSAR PAGE 4.4-37e 4.4.6.5 Instrumentation for Detection of Inadequate Core Cooling The Inadequate Core Cooling Monitoring System installed at Seabrook Station includes the following: Core Exit Thermocouple Monitoring Core Subcooling Margin Monitor Reactor Vessel Level Monitoring The incore thermocouple / core cooling monitor has been installed to provide improved information presentation and display to the plant operators on the status of core heat removal capability. The system monitors all core exit thermocouples and calculates core subcooling margin utilizing redundant channels of instrumentation and Control Room displays. The thermocouple / core cooling monitor utilizes inputs from all core exit thermocouples and Reactor Coolant System wide-range pressure. A microprocessor is employed to calculate saturation temperature from the wide-range pressure and core exit temperature inputs. The Monitoring System displays several-levels of information including: (a) bulk average core exit thermocouple trending; (b) a spatial map exhibiting the thermocouple temperature at its respective location in the core; (c) a core map showing minimum, average, and maximum quadrant temperatures; (d) subcooling margin; (e) a detailed data list exhibiting thermocouple location, tag designation, temperature; and (f) hot channel core exit temperature. The Reactor Vessel Level Instrumentation System (RVLIS) consists of two redundant independent trains that monitor the reactor vessel water levels. Each train provides two vessel level indications:- full range and dynamic head. The full range RVLIS reading provides an indication of reactor vessel water level from the bottom of the vessel to the top of the vessel during natural circulation co;ditions. The dynamic head reading provides an indication of 2eactor core, internals, and outlet nozzle pressure drop for any combination of operating reactor coolant pumps. Comparison of the measured pressure drop with the normal, single phase pressure drop provides an approximate indication of the relative void content of the circulating fluid.

ResDonse to NUREG 0737. II.F.2 " Instrumentation for Detection of Inadequate Core Cooling" The following documentation addresses the requirements of the "Do'cumentation Required" section of NUREG-0137, Section II.F.2. The major section numbers below, identified by Roman numerals, correspond to the equivalent requirement section of the NUREG. 1. Description of System The Inadequate Core Cooling (ICC) Monitoring System installed at Seabrook station includes the following: Core exit thermocouple (T/C) monitoring 1 Core subcooling margin monitor 1 Reactor vessel level monitoring A detailed electrical and layout description of each of the above ICC monitoring subsystems is given below: A. Core Exit Thermocouple System The core exit thermocouple system is a part of the incore instrumentation system. The latter consists of 58 thimble assemblies, each with five fixed neutron detectors and one Type K chromel-alumel thermocouple at fixed core outlet positions. The l core exit thermocouple monitoring system ccasists of two redundant ) --.

independent trains that monitor all 58 of the Seabrook Station chromel-alumel core exit therwecouples (29 on Train 'A' and 29 on Train 'B'). A layout sketch of tha system is shown in Figure 1. The analog indicator on this figure, called " Hot channel'-CET," displays the third highest valid thermocouple temperature. This is consistent with the intent of the WOG Emergency Response Guidelines. The train and quadrant orientation of the thermocouples are presented in Table 1. The core exit thermocouples exit the core through the bottom. After exiting the vessel they are routed for mating with the extension wires (cables) through the seal table where the thermocouple wires are terminated with qualified c,onnectors. Beyond the seal tatle the cables are routed in a manner consistent with the requirements of Regulatory Guide 1.75 to the containment penetrations. The uncompensated core exit thermocouple signals (29) for each train are then sent to their respective ICC Monitors. These monitors house the reference i junction (cold junction) for the thermocouple monitor and thus the net thernal emf of the thermocouple system can be found. The thermocouples have a range of greater than 0 to 2300*F. The thermocouple data is presented in various formats on the plasma displays for quick reference by the operating staff. The display pages contain the folloting information. Core Cooling 1 (Figure 2) - Summary Display - core exit temperature and wide range RCS pressure including presentation ~ of the operational bounds of reactor coolant saturation and the Technical Specification Appendix G, curve; summary of current RVLIS readings. -_

Core Cooling 2 (Figure 3) - Bulk Average CET Temperature Trending - a forty minute trend history of the bulk average CET temperature, including sensor quality data. - = Core Cooling 3 (Figure 4) - Core Exit Thermocouple Map - spatially oriented thermocouple readings, and bulk average CET temperature; one display is provided for each train of thernocouples. Core Cooling 4 (Figure 5) - Core Temperature Map - quadrant core exit thermocouple values (maximum, average, minimum), and subcooling margin. Core Cooling 5 and 6 (Figures 6 and 7) - Core Exit Thermocouple Listing - lists thermocouple location, tag designation, and sensor reading per quadrant. Core Cooling 7 and 8 (Figures 8 and 9) - Train A diagnostic pages with readings status codes for use in equipment servicing. Since these are primarily paintenance-related displays, the control room operating staff would not use these display pages, i B. Core Subcoolina Marain Monitor The inputs, on a train basis, to the core subcooling margin monitor include the following: Wide range RCS pressure (1 channel) Core exit compensated thermocouple values (29 channels).. _ _

The electrical layout of the subcooling margin monitor is shown in Figure 10. The analog indicator shown on this figure displays the subcooling margin. The RCS subcooling margin is calculated based ~ upon the wide range RCS pressure and compensated core! exit thermocouple readings. The value of RCS pressure utilized in the calculation is the output of the data quality algorithm implemented in the subcooling margin monitor. The value of core exit thermocouple temperature is based upon the auctioneered high thermocouple quadrant average temperatures. This averaged value is utilized in the calculation of the core subcooling margin because the quadrant average thermocouple temperature more accurately reflects the individual loop bulk temperature. Basing the core i subcooling margin calculation on the highest themocouple reading would not be indicative of the bulk loop temperatures. Use of the auctioneered high thermocouple quadrant average temperature in the calculation of core subcooling margin is consistent with the utilization in the WOG Emergency Response Guidelines (ERG). The subcooling margin calculated values are routed to both plasma oisplays and analog indicators. The cable routing from sensor input to display meet the requirements of Reg. Guide 1.75. The upper and lower range limits of the core subcooling calculation are adjustable up to a reactor coolant system pressure of 3000 psia. The Regulatory Guide 1.97 limits of 200*F subcooling and 35'F superheat can be accomodated by this calculation. The calculated value, based on auctioneered high quadrant thermocouple input, is displayed on Core Cooling 1, 2, and 4 (Figures 2, 3, and 5, respectively). _ _ _ _.

C. Reactor vessel Level Instrumentatier. System The Reactor Vessel Level Instrum ntation System (RVLIS) consists of two redundant independent trains that monitor the water. level in the reactor vessel. The system provides the operating staff with two instrumentation ranges. The full range RVLIS reading provides an indication of reactor vessel water level f rom the bottom of the vessel to the top of the vessel during natural circulation conditions. The dynamic head RVLIS reading provides an indication of reactor core, internals and outlet nozzle pressure drop for any combination of operating reactor coolant pumps. Comparison of the measured pressure drop with the normal, single phase pressure drop provides an approximate indication of the relative void content of the circulating fluid. The inputs to the RVLIS system on a train basis include the following: 1. Core exit compensated thermocouple values (29 channels) 2. RVLIS capillary line RTD (4 channels) 3. Wide range RCS pressure (1 channel) l 4. Otfferential pressure (2 channels) 5. RVLIS hydraulic isolator contacts (2 channels) A fluid diagram of one train of the Seabrook Station RVLIS system is shown in Figure 11 for the inputs associated solely with the RVLIS system. The electrical block diagram associated with the ~5-

total RVLIS system is shown in Figure 12. The two analog indicators on this figure display the RVLIS readings, one for dynamic head and one for full range. Both trains of RVLIS readings are routed to train oriented plasma displays. The cable routing from sensor input to display meet the requirements of Reg. Guide 1.75. The range of the dynamic head channels corresponds to O to 120% of the delta-p under the reference unvoided. 0% reactor power conditions. The range maximum was selected to bound the cases of operation under conditions dif fering f rom these reference conditions. The range of the full range corresponds to O to 120% of the reactor vessel height. This range will bound all conditions differing from the reference calibration condition. RVLIS readings are presented on the core cooling summary display o (Core Cooling 1, Figure 2) and on the following 3 displays dedicated to RVLIS indications. RVLIS 1 (Figure 13) - RVLIS Dynamic and Full Range Trending - a thirty minute trend history of percent reading for Dynamic Head and present level Full Range Readings, including sensor quality I data tagging. RVLIS 2 (Figure 14) - Reactor Vessel level Instrumentation Fluid System Layout Drawing - provides bulk average CET temperature WR pressure, and subcooling margin. RVLIS 3 (Figure 15) - Train A diagnostic page with readings and status codes for use in equipment servicing. Not used by the control room operating staff. II. Desian Analyses and Test Data Several analyses have been performed to verify the design of the RVLIS system described in Item I.C. The results of these are discussed in the following documents: A. Summary Report, Westinghouse Reactor Vessel level Instrumentation System for Monitoring Inadequate Core Cooling, December 1980 submitted to the NRC via T. M. Anderson to Darrell G. Eisenhut, NS-TMA-2358, dated December 23, 1980, including informal responses to NRC Request for additional information on the Westinghouse RVLIS. 1 B. Supplemental Information on the Westinghouse RVLIS, submitted to the NRC via E. P. Rahe to L. E. Phillips NS-EPR-2519, dated March 19, 1982. In addition to the analyses conducted in the two references above, the hydraulic components of the RVLIS system were installed at the Semiscale Test Facility in Idaho so that transient response characteristics could be obtained during small-break LOCA and other accident conditions. A ~ description of the tests conducted and a discussion of the test results are presented in the following documents:

C. Westinghouse Evaluation of RVLIS Performance at the Semiscale Test Facility, December 1981 submitted to the NRC via E. P. Rahe to L. E. Phillips NS-EPR-2526, dated December 8,1981. D. Westinghouse Evaluation of RVLIS Performance at the Semiscale Test Facility for Test S-UT-8, January 1982 submitted to the NRC via E. P. Rahe to L. E. Phillips, NS-EPR-2542, dated January 13, 1982. E. Westinghouse Evaluation of RVLIS performance at the Semiscale Test Facility for Test S-IB-7 submitted to the NRC via E. P. Rahe to L. E. Phillips, SED-SA-00081, dated June 28, 1982. f Core exit temperature and subcooling margin monitoring rely on direct RCS process neasurement do not involve significantly new design principles. As such, no system testing is necessary. III. Future Testina The acceptability of the RVLIS transient response has been shown in the Idaho semiscale testing. As noted above, core exit temperature and subcooling margin monitoring do not involve significantly new design principles. As such, no additional testing is necessary for any of the 3 ICCI subsystems. s

IV. ResDonse to II.F.2. Attachment I. Desian and Qualification Criteria for Pressurized Water Reactor Incore ThermocouDies A. ThermocouDie location j The Seabrook plant has 58 thernecouples, 29 per train, spread uniformly across the core. Locations are such that during the l l plant life, when power distributions vary from reload to reload, adequate information is provided to the staff to indicate a relative radial distribution of the core exit temperature. The radial distribution is as shown in Figure 4 (Core Cooling 3) for a single train. To determine enthalpy rise, core inlet temperatures j can be approximated by use of wide range cold leg temperature in 1 each loop displayed on the control board. B. Primary Operator DisDlaY CaDabilities O As shown in Figure 4 (Core Cooling 3), a spatially-oriented core map, available on demand, indicating temperature at the core exit, is provided for Train A as the primary operator display. Core exit i temperature at each of the locations can be displayed continuously. In addition, the third highest valid thermocouple temperature is provided on Core Cooling 1 (Figure 2) and on the control board for use with the. plant-specific inadequate core cooling procedures. Trending capability is provided for a period of 40 minutes, as shown in Figure 3, Core Cooling 2. Rapid access to the core exit temperature displays is provided by a single key to the core cooling displays and a page key allowing quick transfer from one display to another. --

C. Backup Display Capability The local display at the ICCM is the fully qualified backup display. The operators have access, via the thumbwheel, switches, 4 to all thermocouple values in engineering units and will be able to access four per quadrant in less than six minutes. The range is i greater than 0 to 2300'F. i D. Tvoes and Locations of Disclavs The display pages are described in Section I, above, and shown in Figures 2-8 for core cooling and Figure 13-15 for RVLIS. The readings displayed by the ICC can be used directly by the j operating staff while implementing the emergency operating procedures. The ' hot channel CET* presented is the third highest thermocouple reading, consistent with the emergency procedures, i The subcooling and RVLIS readings can be used directly to evaluate the respective procedure steps using these indications, i Alarms for core exit thermocouples, RCS subcooling, and RVLIS are l not utilized in the steps of the emergency procedures in order to alert the operating staff. l The ICCI displays are located on Section 8-F (Train 'A') of the Seabrook Station control board and at the Shift Technical Advisor's 1 console (Train '8'). 1 --

E. Design and Qualification for Accident Monitoring Instrumentation i 1. Equipment Qualification Listed below are the appropriate documents indicating the a qualification tests cond0cted on the ICCI s6bsystems. i a. Core Exit Thermocouple Monitoring Subsystem Document 1. Connectors and penetrations NHY EQ File No. t 118-01-01 NHY EQ File No. 4 118-04-01 2. Microprocessors and cabinets ESE-53 l 3. Plasma Display ESE 61A ESE 61B j 4. Control Board Indicators Westinghouse Electric j Corporation F.P. 72789 b. Core Subcooling Margin Monitoring I Subsystem Document 1. Wide Range RCS Pressure ESE-1A 2. Connectors and penetrations NHY EQ File No. 118-01-01 NHY EQ File No. 110-04-01 e, 4 i 4 11

3. Microprocessors ESE-53 4. Plasma Display ESE-61A ESE-61B c. RVLIS Monitoring System [ Subsystem Document 1. Wide Range RCS Pressure ESE-1A 2. Differential Pressure ESE-4 3. Connectors and penetrations NHY EQ File No. 118-01-01 NHY EQ Pile No. 118-04-01 4. Hydraulic Isolator ESE-49 S. Reference leg RTO's ESE-42 6. Microprocessors ESE-53 7. Plas;na Display ESE-63A

8. High Volume Sensor ESE-48 2.

Single Failure Criteria RVLIS, subcooling margin monitors, and inadequate core cooling r monitors are train oriented and therefore meet the single e failure criteria. l 3. Power Supply - The ICC monitors and plasma displays for both trains are powered by 120 VAC vital buses. 4. Channel Availability and Indication I i The operator has access to all ICCI channels at all times pre-and post-accident on the displays discussed in this report. l S. Quality Assurance All hardware associated with the Seabrook Station ICCI monitoring systems meets the applicable portions of the quality assurance regulatory guides. The software of these subsystems is covered by a verification plan that includes review of the associated documentation and testing of the software by verifters independent of the designers. I 6. Direct readout is provided through the plasem graphics. Hard copy capability is provided via data links to the main plant computer system. 7. Capability for Sensor Checks The Seabrook Station ICC Monitor provides the means for quality coding ICCI outputs. Quality codes are displayed on the various plasma display pages. , t

i a 4 8. Capability for Test and Calibration Servicing, testing, and calibration programs are specified to 4 unintain the capability of the monitoring instrumentation. In f addition, a self-testing capability has been provided. This consists of the sensor data checks described above, continuous program diagnostics and self-calibration. Program diagnostics check for system memory failure, mathematical errors, clock and timing errors. Self-calibration is performed serially on each j channel. One channel is taken off line at a time to minimize the impact of the self-test on the ability of the system to s monitor the plant. The self-calibration is initiated every 12 hours. By this self-calibration, the effects of drif t are essentially eliminated in the analog portion of the hardware. 9. Channel Removal from Operation i i i I Whenever means for removing channels from service are included in the design, the design facilitates administrative control of the access of such removal. i

10. Access to Setpoints Adjustments, Calibration and Test Points t

The design facilitates administrative control of the access to I i setpoint adjustments, module calibration adjustments, and test points. i

11. Information Readout The monitoring instrumentation design utilizes human-factored displays to minimize indications potentially confusing to the

~ ~ operator. i I

12. System Repair The instrumentation is designed to facilitate the recognition, l

location, replacement, repair, or adjustment of malfunctioning components or modules.

13. Derivation of System Inputs j

To the extent practicable, monitoring instrumentation inputs are from sensors that directly measure the desired variables. An indirect measurement is made only when it can be shown by l analysis to provide unambiguous information. i

14. Instrumentation Utilization 4

4 To the extent practical, the Seabrook Station ICCI monitoring system has been designed and located in such a manner that the operator uses the ICCI displays during both norwal operation and post accident situations. 1 . i

15. Periodic Testing 1

Periodic checking, testing, and calibration verification are in accordance with the applicable portions of RG 1.118. V. Schedule The Seabrook Station ICCI monitoring system is to be installed and i tested prior to fuel load. Furthermore, the system will be calibrated prior to the plant achieving 5 percent power. i VI. Intearation with Emeroency Doeratina Procedures Seabrook Station is adopting the format and content of the Westinghouse Owners Group (WOG) Emergency Response Guidelines for writing the plant specific procedures. Attachment I illustrates the generic WOG Critical Safety Function Status Tree for monitoring the status of plant core g cooling. As seen, all variables necessary to implement the core 4 cooling status tree are provided by the Seabrook Station ICC instrumentation system. The Functional Restoration Guideline, to which the operator is directed based upon the logic dictated by the tree, also utilizes the information provided by the ICC instrumentation. 1 provides a listing of the generic WOG guideline FR-C.1 ' Response to inadequate Core Cooling.' Note the use of core exit 4 thermocouple temperature in steps 5, 7,16, and 18. Also note that the RVLIS indication is utilized in steps 6, 16, and 23. A review of Seabrook Station procedures FR-C.2, " Response to Degraded Core Cooling," and FR-C.3, " Response to Saturated Core Cooling," also demonstrates the extensive use of ICC instrumentation readings. -

Attachment III provides a listing of the generic WOG guideline E-0, " Reactor Trip'or Safety Injection". Note the use of core exit 4 i thermocodple temperature for calculating RCS subcooling margin in step

25. Similar subcooling margins are utilized throughout the generic guidelines.

VII. Intearation with Control Room Desian Revies The ICCI displays were reviewed as part of the Control Room Design Review. This revieu included a determination of the required characteristics of these displays to support the Emergency Response Procedures and then a verification that the available displays met these requirements. The ICCI displays were found acceptable from this review. d 1 l t j -

TRAIN A TRAIN 8 LOCATION SENSOR LOCATION SENSOR QUAD I 808 TE-35 803 TE-57 C05 TE-33 806 TE-39 C07 TE-28 C08 TE-21 003 TE-42 008 TE-15 F01 TE-51 EOS TE-20 F08 TE-05 F03 TE-30 605 TE-11 F07 TE-07 OUA0 II All TE-54 A09 TE-47 010 TE-23 813 TE-58 014 Te-49 012 TE-31 609 Te-03 E09 TE-12 612 TE-17 E11 TE-21 H11 TE-09 F14 TE-40 HIS Te-43 H13 TE-24 QUA0 III J14 TE-36 JOB TE-01 K12 TE-22 J10 TE-06 LO8 TE-10 L11 TE-18 -o l L10 TE-13 L13 TE-32 i N08 TE-25 L15 TE-52 N14 TE-55 N13 TE-45 i R08 TE-44 P09 TE-37 R11 TE-53 QUA0 IV H03 TE-26 H02 TE-34 J01 TE-46 H04 TE-14 J07 TE-02 H06 TE-04 LOS TE-19 K02 TE-38 NO2 TE-56 K06 TE-08 N06 TE-29 M07 TE-16 i PO4 TE-48 N04 TE-41 I R06 TE-50 I s Table 1. Thermocouple Orientation by Train and Core Quadrant

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U mx m+e 5 = m t O D D D D E"3 C3 O C3 O O O E3 C3 C3 C3 O C3 C3 O O D C3 ru o - = r ru o - = x ru ru ru ~ ~,4 - ~ l 3 I I 'l'l'l'l'l'l'l'l'l'l'l en E 28 E Ov => ~ s g E 7 8, ll m e-W w I or u =5 ru. = w t.o m w w .D-C w C, m, =3 m = l ru .2 = .l.l.l.l.l.l.l.l.l.l.1 g' .4gieii s o C3 C3,o o C C3 C3 a w n.C3 o C C2 e3 W EC C3 SJ1C3 C3 C3 C3 C3 C3 83 C3 C3 C3 E:3 C3 833 C3 03 C3 RJ C3 EIS 653 3*Fu C3EIS 653 2* Fu t4J E1. Li1 L3 Fu Fu e-4 e4 e4 W e-4 EM g C3 0 W

CORE COOLING 3 BULF. R'JG CET XXU Y i 190' E P N M L r. J H G F E D E B R 1 XXXX XXXX 1 2 XXXX 2 3 XXXX XXXX 3 4 XXXX 4 5 XXXX XXXX XXXX 5 E XXXX 5 'l XXXX XXXX 1 90' B XXXX XXXX XXXX XXXX XXXX B 210: 9 XXXX 9 10 XXXX XXXX 10 11 XXXX XXXX XXXX 11 12 XXXX XXXX 12 13 13 19 XXXX XXXX XXXX 19 III II 15 XXXX 15 E P N M L r. J N G F E D E B ' R. Figure 4. Core Cooling 3 g, 39,74 g wennyinis.. Lin.:::c terixvo:"'. all rir. lit.5 esesVNI I l

EDREE00 LING 4 180' i quRa IV l quRa I MRX XXXX T MRX XXXX T RYG XXXX RVG XXXX MIN XXXX l MIN XXXX i SU8 COOL 9De ------------------ g 9 -------------------- 210e I 9URO III l 9URO II MRX XXXX T MRX XXXX T RYG XXXX RYG XXXX MIN XXXX MIN XXXX I a' H Figure 5. Core Cooling 4 (m1

  • id'^ 1" ht y,r.slir,hy,e i le.;rk Oc ;u grh n.,

all sights e.en,Mi

\\ ~ I, i 1 i CORE COOLING 5 i QllADI QUA)II i i LOC SENSOR

  • F LOC SENSOR
  • T l

198 IE-35 M All IE-54 M CBS TE-33 M DIO TE-23 M C07 IE-28 M D14 IE-49 M i. D93 IE-42 M C99 TE-93 M G12 IE-17 M ~ FBI IE-51 M F93 IE-05 M N11 fE-99 M N15 IE-43 M l C95 IE-11 M r i ) I i i 1 Figure 6. Core Cooling 5 cocyg.n'"19 k h3 l' We'.!it'cN5" I *I'*

.3 sinh'r, e-mM

CORE COOLING 6 00ADIII HADIU LOC SENSOR 'F LOC SENSOR

  • F J14 IE-36 M

M93 IE-26 M X12 IE-22 M J01 IE-46 M J07 IE-92 M L98 IE-19 M L18 IE-13 M L95 IE-19 M M98 IE-25 M M82 TE-56 M M14 IE-55 W M96 IE-29 M R98 TE-44 M P94 TE-48 M R11 IE-53 M e Figure 7. Core Cooling 6 c ;rmc'v"' l'.T4 by .b','kt. i.-,..

s-Seabrook Thermocouple Diagnostic Page 1 Train A I' CORE COOLING 7 TIME 09:45:5s MESSAGE 7s l 1 l 1 All s03.379 30 00 G09 s07.023 30 00 LOS s07:591 30 00 l l nos s07.2ss 30 00 G12 60s.012 30 00 L10 s07.902 30 00 l t CDs s05.002 30 00 nos s04.47s 30 00 NO2 607.294 30 00 t [ C07 s08.387 30 00 Mll - s07.294 30 00 Nos s00.208 30 00 l l l D03 608.739 30 00 MIS sos.300 30 00 NOS 600.000 30 00 l l I DIO sos.2ss 30 00 301 s05.251 30 00 N14 s05.973 30 00 l I D14 sos.835 30 00 J07 607.017 30 00 PO4 s08.924 30 00 l l F01 sos.475 30 00 314 sos.374 30 00 m0s 607.940 30 00 l l F08 s07.173 30 00 K12 sos.845 30 00 R11 608.947 30 00 l 1 G0s sos.946 30 00 Los 507.7s3 30 00 1 l I I I l l T/C REFERENCE JUNCTION RTD 70.374 30 00 1 l I 1 I DIAGNOSTIC INFOftMATION I 1 l l l l l i I c.x.g i '4*. ' 1%IS bY V!vi. tin 6.:1 *iw f.k.:i.P 0JINa'"

.;;,....; s. - :. v. < '

Figure 8. Core Cooling 7 i

s-Seabrook Thermocouple Diagnostic Page 2 Train A t - - - - _ _ _ _ _ _ _ _ - - - - - - - - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - ~ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ 1 CORE COOLING 8 TIME 09:44:30 MESSAGE 45 l I i l l MAX I 608.739 30 00 SUBCOOL 44.398 30 00 l 1 AVG I 607.001 30 00 BULK AVERAGE TEMP 606.981 30 00 l } l MIN I 605.002 30 00 MAX QUAD AVERAGE 606.990 30 00 l 1 l MAX II 608.012 30 00 HOT CHANNEL TEMP 608.739 30 00 l I AVG II 606.447 30 00 l l 1 MIN II 603.379 30 00 l l MAX III 608.947 30 00 RC-TI-9424A 608.739 i 1 AVG III 607.422 30 00 RC-TI-9423A +44.398 1 d 1 MIN III 605.973 30 00 l I MAX IV 608.924 30 00 l i 1 AVG IV 606.990 30 00 MIGN INCORE TEMP AIARM STATUS 00 l ! MIN IV 604.476 30 00 SUBCOOL MARGIN AIARM STATUS 00 l 1 1 I I l I I j i DIAGNOSTIC INFORMATION I 1 1 I I I I I cuir.rgi"!?.hI'- 5.'/ce.tinsee iM:ric Conp. :: e'- ali :i;4t., i.tervn' Figure 9. Core Cooling 8 \\ i

I- v 'x hnIe o r dCv i ce J. nc s 1' e t u g e cr u p a g o ' e s l _. o

i. E th.

t t AY m )A ,) f i . A w ^ w(7 s. iwi r dA am t N ws et e ful g Se f a c1 n r As 4 g o I, g a W(t Le a m eS. PD v cn / it. n o a e s 2 N 1 ) !i w I sl 1 I e me il n n* t e s o r u Y^ y r wu m S n

a. e a *'

in pn i on P *' g p oc

o. ro n

oR a i . na r

c. M

_nc o )

m. C

_Ae N' A t s < s mo i ip n c io c o e s e M D ac s< s r* a a* n L A cL a i ~ r t g a n D a ra j il gl l ,I M 8 gn a s t i m eu N l l nva em o m^ e a m. e a o 1 sN c u O b a ar u sr H u v S fo m e a R r u g s a se i n D e k n. s. c o l B QW c o 0 1 / j e r f u / / g i F } C g c / .u /r e T, u 0 t E u i r s f a a 0 m 0 o a s 0 p e 9 e e s 2 E dec e iM C ( ( 'l i:! ji>I i'l i1

i;<

j i'

I:I il j,

ll:\\ ,no i ta to yp l r loa E

h. C v r

E MI .ce i 9r s c 1 tce S s S ic t Et bia e, hig t r - r tP ~ yu l l r, o l C n, a O oh n i t so // l Im C CR A M E-IO LT LT Ii 1' AA AA 1M D (L R RL li t DO l' it YS tK HI YS HI tP i t D uoya 1W L S I L D V T R \\{EO R R @I S H (L N 1 E 1 VS e ru D g T i R F D 2 TR j li mER O i S L ILN HOE N VS s L W E fV

i ,no ti 'r.o Np d I ro e [k c v r ce s 9is r u 8 1

f. c S

W e Fs wM mM i: : r rN 'iUt W m 4 1 t Av8 A h, a g nAA atG aw n si ap gu r I . ten ur e yul t c ) a

ol v

oia m P E D8 p o d 3 i/ C: 1 0 r it m e ~ o !I Il e W l 1 I e e c naY gye r nA 0' age n e Y SL* i s 8 T AP g D u E LS s PE P t C s p S gT ea A t S P f L sso o = E ar = N A eA O C g' sc sf o se N3W I a a K s D Ke M A s u s-A s' u T S n A A T I t A L n D V O R I I I ,l-fo 5 m e s a MA u r rl 0 g Ia 1 3 N a t i a D r ,w kc o l B .o e s" CeS m s UwT u wv ts S S UrC u" s .U Ls 2 u t 'o t s A u A c 1 v A A_T v *" e V Nv Mi A 4 ar e DnN aS a 4 8 Oa 8 P otu r 2 YsO I 2 S 2 fiC 2A so u e P .s ec g i F P p A g co [ i g N = wD c c t f sR / / i A T u T w uE E C r t I h4 mN E n e 8 s e AU 0 s o t c 0 m o 88 u e p a 8 e 8 s s gu s o e n n u n c c ( ( { ,l l !i

l,i!

if!1 ,lif

)1ali!llll'

.o - RVLI5 1 RCP5 RCP5 XDYNAMICHEAD ON % H LL RANGE LN EL OFF BAD -BAD ~ 1 .\\ l i GOOD L-- GOOD 129 120 p. 100 100.... ~. %k. x xf s 89 89 .r 60

G-~\\.

_48, p 40 i tare -3 ptp . antP --- -- 1 RCP ) . gg... >.,.. / i I j i l I I I l I I -30 -20 -10 09 -30 -20 .10 00 MINilIES MINUTES } 1 1 Figure 13. RVLIS 1 { ,j '; :,.f7 h,atm n, [ j ]I all oplit*, res'Vil ~ i i

i RYLIS2 SUBC0.0L FULL RCPS MNANIC RCPS XXX"F RANGE OFF HEAD 0N HRIDI IIRTD2 XXXX XXXX I DP1 DP2 "Ul j Q,) R RfD6 ISO 3 i MLX AYG CET MX "F WRPRESSXXXXPSIG l i i l Figure 14. RVLIS 2 CODYliP.hl' I'~? 0 I"' { av. <.finrhouse' Elect'ic ' o' lier, ion. I all fl;'.hl5 is'58.'fW(! i

Seabrook Rv11s Diagnostic Page Train A l RVLIS 3 TIME 09:46 56 MESSAGE 50 l I I l LT-1311 -1.40s 30 00 FULL RANGE 120.000 34 00 l l LT-1312 24.451 30 00 DYNAMIC HEAD 107.191 34 00 l l l l RC-TE-1313 90.347 30 00 l l RC-TE-1314 92.590 30 00 RC-LI-1311 120.000 l l RC-TE-1318 90.281 30 00 RC-LI-1312 107.191 l l RC-TE-1319 91.603 30 00 l I I l WR PRESS 2230.094 30 00 LIS-1311 00 00 l l LIS-1312 00 00 l l l l l l FULL RANGE IDW LEVEL AIARM STATUS 00 l l DYNAMIC HEAD IDW ALARM STATUS 00 l l SYSTEM TROUBLE ALARM STATUS 00 l l DIAGNOSTIC INFORMATION l I I I I 1 bclpyi r.ht ' P:'.Yb Y W~tinrhor,e'Dectric Co's.aration, .,il tir, hts reserved 1 i Figure 15. RVLIS 3 i F

ATTACHMENT I 0385G:46/BLK/1285

o Number:

Title:

Rev. Issue /Date: F-D.2 CORE COOLING HP/LP, REV.1 1 Sept.,1983 GOTO FR.C.1 4 R C. NO CORE EXIT RVLil NO + TCsLESS FULL RANGE 0 THAN 1200 F GREATER YES THAN (2) YES CORE EXIT 'GOTO 1 0F GOTO FA C.2 AT LEAST RU NI G gygig NO GfEAER YES ~ j THAN (2) ygg RCS GOTO SUBCOOLING NO FA.C.3 8ASED ON i CORE EXIT TC ~ GRE ATER TRAN III (1)'F GOTO FA.C.2 RVLi$ DYNAMIC MEAD RANGE NO GREATER THAN (3)-4 RCP (41-3 ACP (5)-2 RCP YES (6)-1 RCP fA.C.$ CSF SAT

i 1 \\ ATTACHMENT II ) i l j 1 l i S 1 ) D i

A. = pg.C.1 RESPONSE 70 INADEQUATI CORE COOLING HP.Rev. I 1 Sept.1983 d STEP H ACTION /EXPECTID RESPONSE } { RESPONSE NOT CSTAINED b CA UTION

  • If RWSTlevel decreases to less than ni, the SI System should be alignedfor cold leg recirculation using ES-1.3, TRANSFER TO COL.D LEG RECIRCULA TION.
  • Low-head 51 pumps should not be run longer than m without CCWto the RHR heat exchangers.
  • 1 Verify SI Yelve Alignment -

Monvolly olign volves os necessary. PROP!R IM!RGINCY AllGNMENT 2 Verify $1 Fles la AI! Trains: Stort pumps and slign volves os

  • Charging /$i pump flow indicators -

necessary. Try to establish ony CHECK FOR FLOW ether high pressure injection:

  • Migh head 31 pump flow indicators -

(Inter pront specific list]. CHICX FOR FLOW

  • Low head 31 pump flow indicators -

CHECK FOR FLOW ./3 Deck RCP 56,iport Canditions - Try to establish support conditions. AVAILABLE [ Enter plant spevfic list) e f 2 of le l

n e. %, pt.C.1 R15PON5E 70 INADEQUATI CORE COOLING NP Rev. I 1 Sept.1943 l 1 d STEP H ACTION / EXPECTED RESPONSE l l RESPON5E NOT OBTAINED b /4 Oed 51 Acesmerater leerstien Valve 5tstus:

e. Power to isolation volves -
e. Restore power to isoletion volves.

AVAILA8tf

b. Isolation vsfres - OPEN
b. Open isolation volves unless closed efter occumulator discharge.

5 God Cere hit TCs - LESS Go to 5tep 8. TNAN 1200'I e Oed IVL!! Full tange ledication:

s. Indication - GREATER THAN O/
e. lE inuensing. M return to Step 1. g NQ,T, THEN go to Steg,7.
b. Retum to guideline and step in effect 7

God Core hit TCs:

s. Temperature - L115 THAN 700'F
e. g. decreasing THEN return to Step 1. g NQT. THEN go to Step 8.
b. Retum to guideline and step in effect 1

3 of 10

sammeun Itame eso. anse / Deus. FR C.1 RESPONSE 70 INADEQUATE CORI COOLING MP Rev.1 i 5.,t. was d STEP H ACTION / EXPECTED RESPONSE l 5 RESPON5E NOT CBTAINED b ~ NOTE This guideline should be continued while obtaining hydrogen sample in Step 8. s o sk cesteinmen nydrogen concestraties:

s. Obtain a hydrogen concentration moosurement:

[ Enter plant specific means)

b. Hydrogen concentration - LESS
b. Consult plant engineering staff THAN 6.0% IN DRY AIR for additionc! recovery octions.

Go to Step 9.

c. Hydrogen concentration - LESS
c. Turn on hydrogen recombiner THAN 0.5% IN DRY AIR system.

CA UTION

  • Alternate water sourcesfor AFWpumps will be necessary if CST level decreases to less than <<>.

3

  • A faulted or ruptured SG should not be used in h

subsequent steps unless no intact SG is available. 1 Deck Infect SG Levels:

o. Narrow range level-GREATER
e. increcse total feed flow to restore THAN ts/% [r61% FOR ADVERSE narrow ronge level greater hen is/%

a CONTAINMENT) [r6/% for odversy containment). y total feed flow less then (7/ gom. THEN go to Step 18. CBSERVE NOTE ~ PRIOR TO STIP 18.

b. Control feed flow to meintain narrow range level between (s/%

[(6/% for adverse contoinment) and 50% I 4 of 10 1 - ~. -x--,~' ~ ~ ' - ' ' ' ' " ' ' * ' ' * ' ' ' " ^ ' " * * * * ' " ' " * " ' " " " * " " '

l Pt.C.1 RI5PONSI T0 INADEQUATI CORI COOLING NP Rev.1 1 Sept.1933 i d STEP }----{ ACTION /f.XPECTED RESPONSE, [ Af 5PCNSE NOT CSTAINED b 10 Deck RC5 Vest Peths:

e. Power to PRZR PORV block
e. Restore power to block volves.

volves - AVAILABLE

b. PRIR PORVs - CLOSED
b. Monvolly close PRIR PORVs.E eny volve con g be closed, THEN monvolly close its block volve.
c. Block volves - AT LEAST ONE
c. Open block volve unless it was CPEN closed to isolate on open PRZR PORV.

? j

d. Other Rts vent poths CLOSED
d. Cose ony open RCS vent path.

[ Enter plant specifi: Est) NOTE Partial uncovering of SG tubes is acceptable in the i following steps. 11 Depressorize A!!!steet 50s Te (#1 P51G:

e. Dump steem to condenser et
e. Dump steem et monimum rete maximum rate using SG PORVs.
b. Check SG pressures - LESS
b. E 3G pressure decreasing, THEN THAN (#/ PSIG retum to Step 9.g NOT, IfLN, go to Step 18. CBSERVE NOTE PR10e,70 STEP 18.
c. Check RCS het leg temperatures -
c. J! RC5 hot leg temperatures AT LEAST TWO LESS THAN 400'F 4ecreadng, THJ retum to Step 9.E EJ. Ei,N go to 5tep 0 8. 085ERVE NOTE PRIOR 70 STEP.1.
d. Stop 3G depressurization t

l l 5 of 10

i W h Rev. tamus / Deus: i FR.C.1 Rt3PONSI T0 INADEQUATI CORE COOLING NP.Rev.1 1 Sept.1933 d STEP d ACTION / EXPECTED RESPONSE ' 5 RESPONSE NOT CSTAINED }-- 12 Oeek if SI Assamuisters theeld Se isolated:

e. At least two RC5 het leg
o. Go to 5tep 18. CBSERVE NOTE temperatures - LESS THAN PRIOR TO STEP 18.

400*F

b. Oose oil 31 occumulator
b. Vent any unisolated occumulator.

I isolation volves i 13 Step A5 ICPs 14 Depresserfie AR latect SG Te Atmospheric Pressors:

s. Dump steem to condenser of
e. Dump steem of maximum rate maximum rate

. using SG P0RVs. i 15 Verify si Flow: Continue efforts to establish 51 flow. j

  • Chorging/51 pump flow indicators -

Try to establish any other high 3 CHECX FOR FLOW Pressure 'miection: i _og_ [ Enter plant specific Est).

  • High. head 51 pump flow indicators.

E core exit TCs less than 1200'F, CHECX FOR FLOW THEN return to 5tep 14. { NAT. THEN e to Step 18. CSSERVE NOTE, PRIOR -OR-TO STEP 18.

  • Low. head 51 pump flow indicators -

CHECX FOR FLOW s 1 6.f is i

== a. w R.C.1 RESPONSI 70 INADEQUATE CORE COOLING NP.Rev.1 1 Sept.1983 d STEP d ACTION / EXPECTED RESPONSE I [ RESPONSE NOT CBTAINED b 16 Geek Care Caeling

e. Core exit TCs - LESS THAN 1200'F
s. Go to Step 18. DISERVE NOTE PRIOR TO STEP 18.
b. At least two RCS het leg
b. Return to Step 14.

temperatures - LIIS THAN 350'F j ct RVLIS full range indication -

c. Return to Step 14 GREATER THAN (91 l

17 Go To I.1,1055 0F ttACTOR OR SECONDARY COOLANT, Step 12 i NOTE Normal conditions are desired but not requiredfor starting the RCPs. 18 Deck Cers fait TCs. LISS Stan RCPs os necessary until core TNAN 1200'7 exit TCs less then 1200'F. IF core exit TCs greater then 1200'F and oI! evoiichte RCPs ruaning. THEN open oil PRIR PORVs and b!ock volves. ~ g core exit TCs greater then 1200'F ond oil PRZR PORVs and block volves open, THEN open ol! ether RCS vent paths to containment. 7 of le

i \\. M Mg g.C.1 E15PON5170 INADEQUATE CORE COOLING HP Rev.1 i 1 Sept.1933 d STEP H ACTION / EXPECTED RESPONSE ' l RESPONSE NOT CSTAINED b 19 fry To Leestfy Depressertse All Use fovhed or ruptured 50. Inteet SGs Te Atmospheric Pressors:

  • Use PORY

-OR-

  • [ Enter plant specific means) 1 20 God N SI Accenvleters 5hecid Se Isoleted:

\\

e. Low head 51 pump flow
e. Return to Step 13.

Indicators - AT LEAST INTIRMITTENT FLOW

b. Cose off $1 occumulator
b. Vent any unisolated isolation volves occumulator.

1 21 God W RCPs snevid Be Itepped:

e. At least two RC5 hot leg
e. Go to Step 22.

temperatures - LI55 THAN 350'F

b. Stop o!! RCPs l

l 22 Vertfy 51 Plow: Continue efforts to establish 31

  • Chorging/51 pump flow indicators.

flow. Try to establish ony other high CHECX FOR FLOW pressure injection: -OR- [ Enter pront specific list). Migh head $1 pump flow indicators - Return to Step 18. CHECK FOR FLOW -OR-

  • Law. head $l pump flow indicators -

CHECK FOR FLOW j 8 of 10 l

m

m. %,

FR C.1 RISPONSI TO INADEQUATE CORI COOLING MP Rev.1 i 1 Sept.1933 d STEP H ACTION / EXPECTED RESPONSE l { RESPONSE NOT CSTAINED b 23 Oed Core Coeling: Return to Step 18.

  • RVL!$ fw!! range indication -

GREATER THAN tt/

  • At leest two RC5 het leg temperatures - LI55 THAN 350'F 1

24 Se To I.1,1055 0F REACTOR OR SECONDART COOLANT, Step 12 - END-i \\ s l e l 1 ~ e a t 9 ef le

l 1 i l sesamer. Thee. jh a e ; pg.C,1 Rt3P0N5170 INADIQUAft CORI COOLING

  • I 3,,3 FOOTNOTES (1) Enter plant sperfe value corruponding to R WSTswitchover serpoint in plant spreft unns.

(3) Enter plant spectfic time. (3) Enter plant spec $c value whnch it 3-1/2 feet above the bottom of scrivefuelin core with :ero vosd " fraction, plus uncertesntset. Is) Enter plant specWe value terresponding to CSTlow Irvelswitchover setroint in plant specWe unus. (3) Enter plant specWe velut showing SG leve!)nt in the narrow range including allowenessfor normal channelaccurecy. (6) Enter plant specific value showing SG leveljet in the narrow range, including allowancafor normal channel securney, post accident transmitter errors, and rtference les process errors, not to escord 30%. (7) Enter the minimum safeguards ATWpow requirementfor hest removel. pie ellowenenfor normal l channelsecuracy (typically sne MD ATWpump et SG design pressure). (g) Enter plant spec @c value which it 200 psis, minus alloweness for normelchannelsecurecy. t9) Enter plant spectfic value which is above the top of activefW in corr with zero vedfraction, plus uncertaintist. ] l l ~ l 1 I 1 1 t i t i It of it ~ _ _

ATTACHMENT III 9 4 79 0385G:65/8LK/1285

i n.

e
o. :

E.0 REACTOR TRIP OR SAFETY INJECTION i HP R*V 1 i 1 Sept.1983: A. PURPOSE This guideline provides actions to verify proper response of the automatic protection systems following monvol or automatic octuotion of a reactor trip or safety injection, to assess plant conditions, and to identify the oppropriate recovery guideline.

3. $YMPTOMS OR INTRY CONDITIONS
1) The fo!!owing are symptoms that require a reactor trip, if one has not occurred:

[ Enter plant specific setpoints and requirements).

2) The following are symptoms of a reactor trip:
a. Any reoctor trip onnunciator lit.
b. Rapid decreose in neutron level indicated by nuclear instrumentation.
c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.
3) The fo!!owing are symptoms that require a reactor trip and safety injection, if one has not i

occurred: [ Enter plant specific setpoints and requirements).

4) The fo!!owing are symptoms of a reactor trip and safety injections
o. Any $1 onnunciator lit.
b. 51 pumps running.
c. [ Enter plant specific list).

i l 1 of 13 ) .,,...7 _~.-.7

~ m m ~ e .u m, i I0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OSTAINED }-- NOTE

  • Steps 1 through N are IMMEDIA TE A CTION steps.
  • Foldout page should be open.

1 Verify Reacter Trip: Monvally trip reactor. IF reactor will NOT

  • Rod bottom lights - LIT trip, THEN go to FR 5.1, RESPONSE TO
  • Reactor trip and bypass breckers -

NUCLEAR POWER GENERATION / ATWS, 4 OPEN Step 1.

  • Rod position indicators - AT ZERO
  • Neutron flux - DECREASING 2

Verify Ter6ine Trip:

o. All turbine stop volves - CLOSED
o. Manually trip turbine.

b 3 Verify Power Te AC Emergency Besses: l

o. AC emergency busses - AT LEAST
o. Try to restore power to of least one ONE ENERGlZED oc emergen:y bus. IF power con NOT be restored to of least one oc s

emergency bus, THEN go to ECA.O.0, LOSS OF ALL AC POWER, Step 1

b. AC emergency busses - ALL
b. Try to restore power to deencigized ENERGlZED oc emergency busses.

l 4 Check if Si is Actueted: Check if 51 is required. !F.51 is required, [ Enter plant specific means) THEN manually actuate. IF 51 is NOT ( required, THEN go to ES 0.1, REACTOR TRIP RESPONSE, Step 1. i l 1 2 of 13

l m e. %. E0 REACTOR TRIP OR SAFETY INJECTION HP.Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE ! [ RESPONSE NOT OSTAINED W 5 Verify FW isoletion: Manually close volves os necessary.

  • Flow control volves - CLOSED
  • Flow control bypass volves - CLOSED
  • FW isolation volves - CLO,5ED
  • SG blowdown isolation volves -

CLOSED

  • SG sample isolation volves - CLOSED 6

Verify Centoinment isoletion Phase A:

o. Phase A - ACTUATED
o. Manually actuate Phase A.
b. Phase A volves - CLOSED
b. Monvolly close volves.

l j 7 Verify AFW Pumps Rwaning:

o. MD pumps - RUNNING
o. Monvo!!y stort pumps.
b. Turbine driven pump - RUNNING IF
b. Manually open steem supply volves.

NECESSARY l E 8 Verify $1 Pumps Running: Manually stort pumps.

  • Charging /$1 pumps - RUNNING I
  • High head 51 pumps - RUNNING
  • Low head $1 pumps - RUNNING a

'1 9 Verify CCW Pumps - RUNNING Manually stort pumps. ( i l 3 ef13 l l

_= ( = m . w m, I0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 -{ STEP ACTION / EXPECTED RESPONSE l { RESPONSE NOT OBTAINED p 10 Verify Service Water Pumps - RUNNING Monvo!!y stort pumps. 11 Verify Centoinment Ten Ceelers - Monvolly stort fan coolers in emergency RUNNING IN EMERGENCY MODE mode. 12 Verify Centeinment Ventitetion Iseletion:

a. Dompers - CLOSED
o. Monvolly close dampers.

(Appropriate steps for verification of other essential equipment as required by the specific plant design should be placed offer Step 12.] 13 Geck If Main Steemlines $hecid Be Isoleted:

a. [ Enter plant specific means or
o. Go to Step 14.

setpoints]

b. Verify mein steomline isolation and
b. Manually close volves, s

bypass volves - CLOSED 14 Verify Centsineent 5 prey Net Required: i l 2

c. Containment pressure - HAS
o. Perform the following:

REMAINED LESS THAN u) PSIG

1) Verify containment spray initiated.

y!!OT, THEN manually initiate.

2) Verify containment isolation Phase 8 volves closed. H ELT, Ilib monvolly close volves.
3) Stop oil RCPs.

s l s 4 ef13 i l

~ m a. mm. E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l [ RESPON5E NOT OBTAINED i 15 Verify $1 Flow:

o. Chorging/51 pump flow indicators -
c. Manucily stort pumps and align I

CHECX FOR FLOW volves. l

b. RCS pressure - LESS THAN a) PSIG
b. Go to Step 16.

{ [d> PSIG FOR ADVERSE CONTAINMENT)

c. High head 51 pump flow indicators -
c. Monvolly stort pumps and align CHECK FOR FLOW
volves,
d. RCS pressure - LESS THAN (4) PSIG
d. Go to Step 16.

[tJ) PSIG FOR ADVERSE i CONTAINMENT) l

e. Law head 51 pump flow indicarors -
e. Manucily stort pumps and align I

CHECX FOR FLOW volves. g Verify AFW Flow - GREAT!R THAN Manually stort pumps and olign volves os 16 (6) GPM necessary. IF AFW flow greater than (6) gpm con g be established. THEN go to FR H.1, RESPONSE TO LOSS 0F SECONDARY HEAT SINK, Step 1. l 17 Verify AFW Velve A!!gnment - PP.0PER Manually olign volves os necessary. EMERGINCY ALIGNMENT 18 Verify $1 Velve Alignment - PROPER Monvolly align volves os necessary. EMERGINCY ALIGNMENT D 5 of 13 I

I h h Ose. lease /Deana E0 REACTOR TRIP OR SAFETY INJECTION HP R'v 1 1 Sect.1983 d STEP H ACTION / EXPECTED RESPONSE l [ RESPONSE NOT OSTAINED h 19 Deck RC3 Average Tempersters - E temperature.less than (7)*F ond STABLE AT OR TRENDING 70 (7/*F decreasing, THEN: a) Stop dumping steem. b) E,cooldown continues, THEN control total feed flow. Mointain total feed flow greater than (6/ gpm until narrow range level greater than (s;% [(91% for adverse containment] in of least one SG. c) l((cooldown continues, THIN close main steomline isolation and bypass volves. Etemperature greater than (7;'F and increasing, THEN: Dump steam to condenser. -0R-Dump steam using SG PORVs. ~ E 6 ef 13

m m e.mm, E0 REACTOR TRIP QR SAFETY INJECTION HP.Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l { RESPONSE N T OSTAINED }-- 20 Deck PRIR PORY: And $ prey Velves:

a. PORVs - CLOSED
o. IF PRIR pressure less than uo/ psig, THEN monvolly close PORVs. IF ony volve con NOT be closed, THEN j

manually close its block volve. E block volve con M be closed, THEN go to E.1, LOSS OF REACTOR OR i SECONDARY COOLANT, Step 1. i

b. Normal PRIR spray volves - CLOSED
b. E PRZR pressure less than n/> psig, l

THEN monuelly close volves. E volves con NOT be closed, THEN stop RCP(s) supplying failed spray volve(s). NOTE Sealinjection flow should be maintained to all RCPs. i' r 21 Geck if RCPs $heeld Be Stepped: j

o. Si pumps - AT LEAST ONE RUNNING
o. Go to Step 22.

l

  • Chorging/51 l

-OR-

  • High. head Si
b. RCP Trip Parameter - LESS THAN In/
b. Go to Step 22.

[UJ/ FOR ADVERSE CONTAINMENT]

c. Stop oil RCPs

{ l l 1 l 7 of 13 4 i , _.. _, _ _ _ _,. _ ~..,, - -.., _ - -,,_,-_,.-,__.-.__,,..,.., _ _...- - - -,,..--.,

W h gov,see stes : I.0 REACTOR TRIP OR SAFETY INJECTION NP.Rev. I 1 Sept.1983 I STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED h 22 Check if SGs Are Not Feelted:

c. Check pressures in oil SGs -
c. Go to E 2, FAULTED STEAM
  • NO SG PRESSURE DECREASING IN GENERATOR ISOLATION, Step 1.

AN UNCONTROLLED MANNER

  • NO SG COMPLETELY DEPRE55URIZED 23 Check if SG Tubes Are Not Ruptured:

Go to E 3,5 TEAM GENERATOR TUBE i

  • Condenser air ejector radiation -

RUPTURE, Step 1. NORMAL

  • SG blowdown radiation - NORMAL 24 Check if RC5 is intact:

Go to E 1, LOSS OF REACTOR OR

  • Containment radiation - NORMAL SECONDARY COOLANT, Step 1.

r

  • Containment pressure - NORMAL

(,

  • Containment recirculation sump level

- NORMAL S 1 i d i 1 8 ef 13

m nm E.0 REACTOR TRIP OR SAFETY INJECTION HP.Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OST/ 'NED h 25 Check if St Flow Should Be Reduced:

a. RCS subcooling based on core exit
a. DO NOT STOP 51 PUMPS. Go to TCs - GREATER THAN (14/*F Step 27.
b. Secondary heat sinks
b. jf neither condition satisfied, THEN
  • Total feed flow to SGs - GREATER DO NOT STOP 51 PUMPS. Go to THAN (6/ GPM Step 27.

-OR-

  • Narrow range levelin at least one SG - GREATER THAN (#/%
c. RCS pressure - STABLE OR
c. DO NOT STOP 51 PUMPS. Go to INCREASING Step 27.
d. PRZR level - GREATER THAN //J/%
d. DO NOT STOP 51 PUMPS. Try to i

stabilize RC5 pressure with normal e ( PRZR spray. Return to Step 250. 26 Go To E51.1,51 TERMINATION, Step 1 l (~ i 4 9 of 13 1 ,,w, er- ,n ,.m.--

l m, n in a.m. E.0 REACTOR TRIP OR SAFETY INJECTION HP.Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l l RESPONSE NOT OSTAINED F 27 laitiet. Monitoring of Cruical Sefety Function Stetus Trees CA UTION Alternate water sourcesfor AFWpumps will be necessary if CST level decreases to less than aci. 28 Geek $G Levels:

a. Narrow range level - GREATER THAN
o. Maintain total feed flow greater than

(#>is (61 gpm until narrow range level greater than (s>% in of least one SG.

b. Control feed flow to maintain narrow
b. IF narrow range levelin any SG range level between (s/?& ond 50%

continues to increase in on urcontrolled manner, THEN go to E 3, [ SIEAM GENERATOR TUBE RUPTURE, ( 5tep 1. 2g Geek Secondary Radiation - NORMAL Go to E 3, STEAM GENERATOR TUBE (Enter plant specific means) RUPTURE. Step 1. 30 Geck Auxiliary Building Rodiet~tn - Evoluote cause of abnormal conditions. IF, NORMAL the cause is a loss of RCS inventory outside containment, THEN go to ECA 1.2. LOCA OUTSIDE CONTAINMENT, Step 1. f 31 Deck PRT Conditions - NORMAL Evoluote cause of obnormal conditions. 10 of 13 l t

feesdsort fWes ass. Issuehs E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE ' l RESPONSE NOT OBTAINED F i CA UTION If offsite power is lost after SI reset, manual action may be required to restart safeguards equipment. 32 Roset$1 33 Reset Containment isoletion Phose A And Phase B 34 Establish lastrument Air To Centeinment Start one air compressor and establish instrument air to containment. CA UTION RCS pressure should be monitored. If RCS pressure decreases to less than u> psig the low-head SI pumps must be manually restarted to supply water to the RCS. 35 Check if Low.Need 51 Pumps should Be Stepped:

c. Check RCS pressure:
1) Pressure - GREATER THAN
1) Go to E-1, LOSS OF REACTOR OR g) PSIG SECONDARY COOLANT, Step 1.
2) Pressure - STABLE OR INCREASING
2) Go to Step 36.
b. Stop low head 51 pumps and place in stondby 11 of 13 l

Mweben TWies Sov.leawe/Deses E.0 REACTOR TRIP OR SAFETY INJECTION HP.Rev. I 1 Sept.1983 1 -{ STEP d ACTION / EXPECTED RESPONSE ' l RESPONSE'NDT OBTAINED 36 Deck if Diesel Generstors should Be Stepped:

o. Verify oc emergency busses -
c. Try to restore offsite power to oc ENERGlZED BY OFFSITE POWER emergency busses. E offsite power j

con NOT be restored. THEN load the folloEg equipment on oc emergency i i busses: [ Enter plant specific list).

b. Stop any unloaded diesel generator and place in standby 37 Return To Step 19

- END - r-1 1 ( 12 of 13

mn m s. ,o E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 FOOTNOTES ~. (1) Enter plant specy'c containment pressure serpointfor spray actuation. i (2) Enter plant specific valuefor the shutoff head prenure of the high head Sipumps, plus allowances for normal channelaccuracy. (3) Enter plant specipe value for the shutoff head pressure of the high-head Sipumps, plus allowances for normalchannelaccuracy andpost accident transmitter errors, not to exceed 2000 psig. (i) Enter plant specipe valuefor the shutoff head pressure of the low head Sipumps, plus allowancesfor normal channelaccuracy. (3) Enter plant specyic valuefor the shutoff head pressure of the low head 51 pumps, plus allowancesfor normal channel accuracy and post accident transmitter errors. (6) Enter the minimum safeguards AFWflow requirementfor heat removal, plus allowancesfor normal channe! accuracy (typically one MD AFWpump capacity at SG design prusure). (7) Enter plant specific no-load temperature. (8) Enter plant specific value showing SG leveljust in the narrow range, including allowancesfor normal channel accuracy. (9) Enter plant specific value showing SG leveljust in the narrow range, including allowances for normal channelaccuracy, post accident transmitter errors, and reference leg procus errors, not to exceed 30*'s. 110) Enter PRZR POR Vprusure serpoint. (11) Enter PRZR Spray prusure serpoint. (12) Enter plant specific RCP trip parameter and serpoint, including allowances for normal channel ac-curacy. Refer to document RCP TRIP /RESTARTin Generic issues section of Executive Volume. (13) Enter plant specific RCP trip parameter and setpoint, including allowancesfor normal channel accuracy and post accident transmitter errors. Refer to document RCP TRIP!RESTARTin Generic luurs sec. tion of Erecutive Volume. (14) Enter sum of temperature and pressure measurement system errors, including allowancesfor normal channelaccuracies, translated mto temperature using saturation tables. !!!) Enter plant specific value showing PRZR leveljust in range, including allowances for normal channel accuracy. (16) Enter plar.t specific value corresponding to CSTlow levelswitchover serpomt in plant specific units. 13 of 13 -}}

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