Text

Forwards Response to NUREG-0737,TMI Item II.F.2 & SER Confirmatory Item 15 Re Inadequate Core Cooling Monitoring Sys.Subset of Monitoring Sys That Provides Rev 2 to Reg Guide 1.97 post-accident Instrumentation Consolidated
	ML20133K772
Person / Time
Site:	Vogtle
Issue date:	08/05/1985
From:	Bailey J; GEORGIA POWER CO.
To:	Adensam E; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097, TASK-2.F.2, TASK-TM GN-674, NUDOCS 8508120366
	Download: ML20133K772 (71)
	v • d • e

Georgia Power Comp' ry Route 2 a

Waynes. Box 2994 boro, Georgia 30830 Telephone 404 554 9961 404 724 8114 Southern Company Services. inc.

Post Office Box 2625 Birmingham, AktDama 35202 Telephone 205 970-so11 b

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August 5, 1985 Director of Nuclear Reactor Regul ti Attention:

Licensing Branch #4Ms. Elinor G. Adensam, Chief a on File: X7BC35 Division of Licensing Log:

GN-674 Washington, D.C.U.S. Nuclear Regulatory Commissio n

20555

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CONSTRUCTION PERMIT NUMBERS CP I

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V6GTLE ELECTRIC GENERATING PLANT AND CPPR-1209 i

iSER CONFIRMATORY ITEM-15:

- UNITS 1 AND 2

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Dear Mr. Denton:

TMI ITEM-II.F.2 i

is a subset of the Plant Saf tThe Inadequate Core Cooling (ICC

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i e y Monitoring System (PSMS) that ponitorin Regulatory Guide 1.97 Rev.

consolidation.

2 post accident monitoring instru ogtle As discussed in the attached respo 1

rovides Instrumentation for Detection of I

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mentation display the post accident monitoring i i

nadequate Core Cooling, the PSMS displnse to panel displays.

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nstrumentation on seismically qualifi d The preliminary ICC monitoring di Attachment I of the attached resp

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ays the latest generation Westingh e flat splays are included in onse.

j These displays are characteristic the Reactor Vessel Level Instr ouse designed ICC monitoring hardwar j

to be installed at several NTOL' of umentation System (RVLIS), installed oe, including s.

Please note that portions of thi r planned in order toBy copy of this letter, Westingh ouse will release the NRC from thes transm make a microfilm i

j extent of the copyright release erial.

copy necessary for their records copyright previously published in Appendix 4A The non-copyrighted material has bThis is the of the VEGP FSAR.

If your staff requires any additi 1.

een contact me.

onal information, please do not hesi Sincerely, tate to o

J. A. Bailey Project Licensing Manager j

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g8120366 E

ADOCK 850805 050004p4 PDR

t Georgia Powtr Comhey

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Rout 3 2, Box 299A '

Waynesboro, Georgia 30830 Telephone 404 5549961 404 724 8114 Southern Company Services,Inc.

Post Office Box 2625 Birmingham, Amoama 35NI.-

Telephone 205 970W11 Vogtle Proj,ect g

qy August 5, 1985 Director of Nuclear Reactor Regulation File: X7BC35 Attention:

Ms. Elinor G. Adensam, Chief Log:

GN-674

' Licensing Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555 NRC DOCKET NUMBERS 50-424 AND 50-425 C0ljSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-1209 i

VdQTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 l

' SER CONFIRMATORY ITEM-15: TMI ITEM-II.F.2 j

Dear Mr. Denton:

I s

The Inadequate Core Cooling (ICC) Monitoring System installed on Plant Vogtle is a subset of the Plant Safety Monitoring System (PSMS) that provides Regulatory Guide 1.97 Rev. 2 post accident monitoring instrumentation display consolidation. As discussed in the attached response to NUREG-0737, II.F.2, l

Instrumentation for Detection of Inadequate Core Cooling, the PSMS displays the post accident monitoring instrumentation on seismically qualified flat

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panel displays. The preliminary ICC monitoring displays are included in Attachment I of the attached response. These displays are characteristic of the latest generation Westinghouse designed ICC monitoring hardware, including the Reactor Vessel Level Instrumentation System (RVLIS), installed or planned 7

to be installed at several NTOL's.

Please note that portions of this transmittal contains copyrighted material.

By copy of this letter, Westinghouse will release the NRC from the copyright in order to make a microfilm copy necessary for their records. This is the extent of the copyright release. The non-copyrighted material has been

]

previously published in Appendix 4A of the VEGP FSAR.

If your staff requires any additional information, please do not hesitate to d

contact me.

t Sincerely, m

o e

i J. A. Bailey l

Ul Project Licensing'4anager b hl r[:;

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JAB /caa Attach:nent xc:

D. O.-Foster G. Bockhold, Jr R. A. Thomas T. Johnson (W/o Att.)

J. E. Joiner, Esquire D. C. Teper (W/o Att.)

B. W. Churchill, Esquire L. Fowler M. A. Miller W. C. Ramsey B. Jones, Esquire (W/o Att.)

Vogtle Project File L. T. Gucwa 0053V

Response to NUREG-0737. II.F.2

" Instrumentation for Detection of Inadequate Core Cooling" I.

The Inadequate Core Cooling (ICC) Monitoring System installed at Plant Vogtle Project will include the following:

Core exit thermocouple (T/C) monitoring Core subcooling margin monitoring 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 monitoring system consists of two redundant independent trains that monitor all 50 of the Plant Vogtle chromel-alumel core exit thermocouples (25 on protection set III and 25 on protection set IV). A layout sketch of the system is shown in Figure I.

The core exit thermocouples are mounted at the top of core support plate. They are then routed to four upper head conoseal penetrations.

After exiting the conoseal penetrations, the thermocouple wires proceed through a swaglok and then to qualified connectors to facilitate disconnection during removal of the upper head.

Upon exiting the reactor vessel 0290G:2/ GEL /4-85

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cavity, the cables are routed in a manner consistent with the i

requirements of Regulatory Guide 1.75 to the in-containment qualified reference junction boxes. Each reference junction box includes three redundant platinum RTD's imbedded in a block of copper to reflect the temperature at the junction of the chromel alumel and copper wire. The uncompensated core exit thermocouple signals (25) and the reference junction box temperatures (3) are routed to Remote Processing Units (RPU) A3 and B3.

The signals from both RPU's are routed to both Display Processing Units (DPU) for calculation of the compensated core exit thermocouple value.

The value chosen for the reference junction box temperature is a function of the data quality of each of the RTD signals.

Following the calculation of all 50 compensated thermocouple values, the information from both DPU's are transmitted to both seismically qualified flat panel Plant Safety Monitoring System (PSMS) displays. The displays are located on Section D of the Plant Vogtle control board as shown in the Vogtle FSAR, Figure 18.1-2.

DPUA and display A are powered by train A and DPUB and display B are powered by train B.

The cabling between the RPU's, i

DPU's and displays meet the requirements of Regulatory Guide 1.75.

l B.

Core Subcoolina Marcin Monitor l

i i

The inputs to the core subcooling margin monitor include the l

l following:

Wide range RCS pressure (4 channels)

Core exit compensated thermocouple values (50 channels) r Reference junction box RTD values (6 channels) 0290G:3/ GEL /4-85

The electrical layout of the subcooling margin monitor is shown in Figure II. One channel of wide range RCS pressure is input into each RPU channel (A2, A3, B2, and B3). Also 25 uncompensated thermocouple channels and the corresponding 3 reference junction box RTD signals are input into RPU's A3 and B3.

The outputs of each of the RPU's are routed to each DPU. The RCS subcooling margin is then calculated based upon the wide range RCS pressure and compensated core exit thermocouple readings. The value of RCS pressure utilized in the calculation is a function of the data quality of the pressure readings. The value of core exit thermocouple temperature is based upon the auctioneered high thermocouple quadrant average temperatures. The auctioneered high thermocouple quadrant average temperature is utilized in the calculation of the core subcooling margin for the quadrant average thermocouple temperature more accurately reflects the individual loop bulk temperature.

Basing the core subcooling margin calculation on the highest thermocouple 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 WOG ERG's do not specify that core subcooling margin shall be based upon maximum core exit temperature. The WOG ERG's do specify that the core cooling status tree utilize the fifth hottest core exit thermocouple temperature in the implementation of the decision paths, however, core subcooling margin may be calculated using average core exit thermocouple temperatures. The subcooling 0290G:4/ GEL /4-85

margin calculated values are routed to both displays (A and B).

The cable routing from sensor input to display meet the requirements of Reg. Guide 1.75.

The PSMS displays are the same display panels utilized in displaying tta core exit thermocouple information.

C.

Reactor Vessel Level Instrumentation System The Reactor Vessel Water Level System (RVWL) consists of two redundant independent trains that monitor the water level in the reactor vessel.

The wide 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 conditions. The narrow range RVLIS reading provides an indication of reactor vessel water level from the middle of the hot leg pipe to the top of the reactor vessel head 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 nornal, single phase pressure drop provides an approxinate indication of the relative void content of the circulating fluid. The inputs to the RVLIS system include the following:

0290G:5/ GEL /4-85

1.

Core exit uncompensated thermocouple values (50 channels) 2.

Reference junction box RTD values (6 channels) 3.

Wide range RCS pressure (4 channels) 4.

Differential pressure (6 channels) 5.

Reference leg temperature values (14 channels) 6.

Reactor coolant pump status (4 channels)

A fluid diagram of one train of tb;- Plant Vogtle RVLIS system is shown in Figure III for the inputs associated solely with the RVLIS system. The electrical ') lock diagram associated with the RVLIS system is shown in Figure IV.

As discussed, the core exit thermocoupla readings and reference junction box temperatures are input to RPU's A3 and B3. Also, one wide range RCS pressure channel is input into each RPU (A2, A3, B2, and B3).

)

In addition, one of two sets of three differential pressure signals (wide range, narrow range, and dynamic head) are input into RPU A3 and B3, respectively. Also seven reference leg compensating temperature inputs from each train of RVLIS are input into RPU's A3 and B3.

Finally, to determine the appropriate RVLIS indication, the running status of each reactor coolant pump is input into the non-1E RPU N1.

0290G:6/ GEL /4-85

Both trains of RVLIS readings are routed to both displays (A and j

B). The cable routing from sensor input to display meet the j

requirements of Reg. Guide 1.75.

The PSMS displays are the same j

display panels utilized in displaying the core subcooling margin and the core exit thermocouple information.

II.

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.

B.

Responses to NRC Request for Additional Information on the Westinghouse RVLIS, Summary Report.

C.

Supplemental Information on the Westinghouse RVLIS, submitted to the NRC via E. P. Rahe to L. E. Phillips, NS-EPR-2579 dated March 19, 1982.

In addition to the analyses conducted in the three 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 i

accident conditions.

A description of the tests conducted and a discussion of the test results are presented in the following documents:

0290G:7/ GEL /4-BS

D.

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.

4 E.

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.

F.

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.

III.

A description of the tests conducted on the Westinghouse RVLIS system and the results of the tests are presented in references (D), (E), and (F) listed above.

I IV.

Response to II.F.2, Attachment I, Design and Qualification Criteria for Pressurized Water Reactor Incore Thermocouples A.

Attachment I to this response, provides the preliminary design of the display package on the PSNS. The display package hierarchy, as summarized in Exhibit 1, includes the following:

1.

Top Level Plant Status Summary (Exhibit C-2.0) 0290G:8/ GEL /4-85

7 _

2.

Four Lower Level Graphic Displays a.

Core Temperature Map (Exhibit C.3-1) b.

Pressure-Temperature Operating Limits (Exhibit C-400) c.

Reactor Vessel Water Level (Exhibit C-5.0) d.

Nuclear Power (Exhibit C-6.0) 3.

Four Pages of Menu Display a.

Primary Data Trend Menu b.

Secondary Data Trend Menu c.

Containment Data Trend Menu d.

Detailed Data Menu 4.

Four Multi-Page Sets of Data a.

Six Page Set of Primary Data Trends b.

Five Page Set of Secondary Data Trends 0290G:9/ GEL /4-85

c.

Two Page Set of Containment Data Trends d.

Eight Page Set of Detailed Data B.

The following exhibits provide a top down display of the core exit thermocouple information.

1.

a.

Exhibit C-2.0 - maximum core exit thermocouple temperature, b.

Exhibit C-3.1 - quadrant core exit thermocouple maximum, average and minimum temperature. Also provides a comparison between the RCS hot leg RTD's and the quadrant T/C data.

c.

Exhibit C-10.6A - spatially oriented core exit thermocouple map showing each thermocouple temperature.

d.

Exhibit C-10.4 and C-10.5 - Alpha numeric listing of core exit thermocouple location, tag designation and temperature reading per quadrant.

e.

Exhibit C-7.3A - a two hour trend history of the three core exit thermocouple quadrant maximum temperatures.

0290G:10/ GEL /4-85

C.

The following exhibits provide a top down display of the core subcooling margin (based upon core exit thermocouples):

i 1.

a.

Exhibit C-2.0 - core subcooling margin based upon core exit thermocouples.

b.

Exhibit C-4.0 - RCS pressure - temperature plot exhibiting plant approach to saturation, c.

Exhibit C-10.1 - alpha numeric listing of both trains of core subcooling margin.

d.

Exhibit C-7.1 - a two hour trend history of the core subcooling margin.

D.

The following exhibits provide a top down display of the RVLIS system.

1.

a.

Exhibit C-2.0 - displays appropriate RVLIS narrow and wide range and dynamic head readings depending upon RCP

status, b.

Exhibit C-5.0 - mimic of analog meters indicating RVLIS narrow, wide and dynamic readings with respect to reactor vessel. Only displays appropriate ranges based upon RCP status.

0290G:11/ GEL /4-85

c.

Exhibit C-10.2 and C-10.3 - alpha numeric listing of appropriate ranges for both trains of RVLIS system.

d.

Exhibit C-7.6 - a two hour trend history of all three RVLIS ranges. Also presents a trend of RCP status.

E.

Trend Capability - In addition to being displayed on the PSMS, the RVLIS readings are recorded on the main control board.

Furthermore, the core subcooling margin, core exit thermocouple temperature and the RVLIS indications are trended on the Vogtle Safety Parameter Display System.

F.

Alarm Capability - The core exit thermocouple display pages are designed such that any numeric thermocouple readout greater than 1200*F will be displayed in inverse video and flashed at a frequency of 1 hertz.

The core subcooling margin will indicate "SUBC00L" when the auctioneered high quadrant thermocouple average temperature is at or below the RCS coolant saturation point.

"SUBC00L" and the respective numeric value in degrees F will be displayed in inverse video when the subcooling margin is less than a specified value.

"SUPERHEAT" and the respective numeric value in degrees F will be displayed in inverse video and flashed at a frequency of 1 hertz when the auctioneered high quadrant thermocouple average temperature exceeds the coolant saturation temperature.

0290G:12/ GEL /4-85

G.

Backup Display - Since the Plant Vogtle PSMS display system features two redundant independent displays, one display console is considered the primary display and the other display console is considered the backup display. As such, the backup display console for ICC monitoring is also a qualified display.

H.

Location - The PSMS displays are located on Section 0 of the Plant Vogtle control board as shown in Figure 18.1-2 of the FSAR.

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U $.N V.

Response to II.F.2, Appendix B, Design and Qualification Criteria for Accident Monitoring Instrumentation A.

Equipment Qualification 1.

Core Exit Thermocouple Monitoring Listed below are the appropriate documents indicating the qualification tests conducted on the PSMS subsystems.

Subsystem Document a.

T/C Connectors and Adaptors ESE-43B,C b.

Reference Junction Box ESE-44A c.

Microprocessors ESE-53 d.

Plasma Display ESE-61B 0290G:13/ GEL /4-85

Insert IV.H The Plant Vogtle control room design review (CRDR) includes a task analysis, based on the Westinghouse owner group (WOG) Emergency Response Guidelines (ERGS). This analysis will include a review of needed instrumentation and required instrument characteristics to support the operator tasks in the Emergency Operating Procedures (EOPs). The ICCI par-ameters in the WOG ERG's are incorporated in Vogtle specific E0P's and the use of these procedures is included in the operator training program. The priority of critical safety functions in the WOG ERG's are incorporated in the Vogtle E0P's and associated alarms.

2.

Core Subcooling Margin Monitoring Subsystem Document a.

Wide Range RCS Pressure ESE-2 b.

Core Exit Thermocouples See Item Above c.

Microprocessors ESE-53 d.

Plasma Display ESE-61B 3.

RVLIS Monitoring System Subsystem Document a.

Wide Range RCS Pressure ESE-1A b.

Differential Pressure ESE-4 c.

Core Exit Thermocouples See Item Above d.

High Volume Pressure Sensor ESE-48 e.

Hydraulic Isolator ESE-49 f.

Reference Leg RTD's ESE-42 0290G:14/ GEL /4-85

g.

Microprocessors ESE-53 h.

Plasma Display ESE-61 B B.

Single Failure Criteria A detailed discussion of the Reg. Guide 1.97 Post Accident Monitoring Design Basis is presented in Section 7.5 of the Plant Vogtle FSAR.

Included in the discussion is a justification for the number of channels selected and the diverse variable identified where necessary.

Presented in the Vogtle FSAR, Section 7.5, Table 7.5.2-1 is a detailed description of the characteristics associated with each ICC monitoring system, including range, number of channels, and qualification status.

C.

Power Supply - RPU's Al and A2, DPUA and Display A are powered by inverter power bus I.

RPU's B1 and B2, DPUB and display B are powered by inverter power bus II.

RPU A3 is powered by inverter power bus III and RPU B3 is powered by inverter power bus IV.

A sketch of signal flows between the protection channels, RPU's, DPU's, and displays is shown in Figure V.

D.

Channel Availability and Indication The operator has access to all ICCI channels at all times pre-and post-accident on several QDPS displays.

These include the Exhibit 0290G:15/ GEL /4-85

e C-2.0 display, and the detailed data display of Exhibits C-10.1, C-10.2, C-10.3, C-10.4, C-10.5, and C-10.6A.

The recording capability of the ICCI channels is indicated in the Plant Vogtle FSAR Table 7.5.2-1.

E.

Quality Assurance All hardware associated with the Plant Vogtle PSMS ICCI monitoring systems meets the applicable portions of the quality assurance regulatory guides.

F.

Capability for Sensor Checks The Plant Vogtle PSMS provides the means for cross checking between channels that bear a known relationship to each other.

In addition, the subsystem displays only project a group value based upon a data quality algorithm.

Quality codes that may be displayed include GOOD, POOR, BAD and SUSPECT. The operator may access the lower level detailed data lists to determine the reason for other than GOOD data quality group values.

G.

Capability for Test and Calibration See Plant Vogtle FSAR, Section 7.5.2.3.1.3D.

H.

Channel Removal from Operation See Plant Vogtle FSAR, Section 7.5.2.3.1.3E.

0290G:16/ GEL /4-85

I.

Access to Setpoints Adjustments, Calibration and Test Points See Plant Vogtle FSAR, Section 7.5.2.3.1.3F.

J.

Information Readout See Plant Vogtle FSAR, Section 7.5.2.3.1.3G.

K.

System Repair See Plant Vogtle FSAR, Section 7.5.2.3.1.3H.

L.

Derivation of System Inputs See Plant Vogtle FSAR, Section 7.5.2.3.1.31.

M.

Instrumentation Utilization To the extent practical, the Plant Vogtle PSMS display has been designed and located in such a manner that the operator uses the ICCI displays during both normal operation and post accident situations.

N.

Periodic Testing See Plant Vogtle FSAR, Section 7.5.2.3.1.33.

0290G:17/ GEL /4-85

VI.

Schedule The Plant Vogtle ICCI monitoring system is to be installed, tested and calibrated prior to fuel load.

VII.

Plant Vogtle is adopting the format and content of the Westinghouse Owners Group (WOG) Emergency Response Guidelines for writing the plant specific procedures. Attachment II illustrates the generic WOG Critical Safety Function Status Tree for monitoring the status of plant core cooling. As seen, all variables necessary to implement the core cooling status tree are provided by the Plant Vogtle 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.

Attachment III provides a listing of the generic WOG guideline FR-C.1

" Response to Inadequate Core Cooling." Note the use of core exit 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 Plant Vogtle 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 IV provides a listing of the generic WOG guideline E-0,

" Reactor Trip or Safety Injection".

Note the use of core exit thermocouple temperature for calculating RCS subcooling margin in step 25.

Similar subcooling margins are utilized througho9t the generic guidelines.

0290G:18/ GEL /4-85

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ATTACHMENT II 0290G/ GEL /1-85

Numb:r.

Titis:

Rsv. Issu2/Dat::

F-0.2 CORE COOLING HP/LP, REV.1 1 Sept.,1983 GO TO FR-C.1 1

?R C.S CORE EXIT RVLIS NO TCsLESS FULL RANGE 0

THAN 1200 F GREATER YES THAN (2)

CORE EXIT GO TO TCs LESS FR C.2 0

THAN 700 F YES GO TO FR-C.2 NO AT LEAST ONE RCP RUNNING RVLIS NO FULL RANGE YES GREATER THAN (2)

YES GO TO COOLING NO FR C.3 BASED ON CORE EXIT TCs Ofop EATER TRAN YES j

GO TO FR-C.2 RVLIS DYNAMIC HEAD RANGE NO GREATER THAN (3)-4 RCP (4)-3 RCP (5)-2 RCP YES (6)-1 RCP

?R C.$

CSF

'q SAT F

v

- ~

v ATTACHMENT III 0290G/ GEL /1-85

\\

Musnhor:

TMes Rev. ineue/Dese:

FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE l[

RESPONSE NOT OBTAINED CA UTION

If RWST level decreases to less than ni, the SI System should be alignedfor cold leg recirculation using ES-1.3, TRANSFER TO COLD LEG RECIRCULA TION.

Low-head SIpumps should not be run longer than a> without CCW to the RHR heat exchangers.

1 Verify si Volve Alignment -

Manually align valves as necessary.

PROPER EMERGENCY ALIGNMENT 2

Verify $1 Flow in All Trains:

Start pumps and align valves as

Charging /Si pump flow indicators -

necessary. Try to establish any CHECK FOR FLOW other high pressure injection:

High-head Si pump flow indicators -

[ Enter plant specific list].

j CHECK FOR FLOW Low-head Si pump flow indicators -

CHECK FOR FLOW

/3 Check RCP Support Conditions -

Try to establish support conditions.

AVAILABLE (Enter plant specific list]

1 i

2 of 10

w THies Rev. laswosDeses FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev.1 1 Sept.1983 l

i ACTION / EXPECTED RESPONSE '

RESPONSE NOT CBTAINED STEP l

/4 Check 51 Accumulator Isolation Volve Status:

a. Power to isolation valves -

o. Restore power to isolation volves.

AVAILABLE

b. Isolation volves - OPEN

b. Open isolation valves unless closed offer occumulator discharge.

5 Check Core Exit TCs - LESS Go to Step 8.

THAN 1200*F 6

Check RVLIS Full Range Indication:

a. Indication - GREATER THAN (3)

a. E increasing, THEN return to Step 1. g NOT, THEN go to Step,7.

b. Return to guideline and step in effect 7

Check Core Exit TCs:

a. Temperature - LESS THAN 700*F

o. E decreasing THEN return to Step 1. E NOT, THEN go to Step 8.

b. Return to guideline and step in effect 3 of 10

Musniner TH6es now.isme/Deve:

FR C*1 RESPONSE TO INADEQUATE CORE COOLING HP-Rev.1 1 Sept.1983 ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED STEP NOTE This guideline should be continued while obtaining hydrogen sample in Step 8.

8 Check Containment Hydrogen Concentration:

a. Obtain a hydrogen concentration measurement:

[ Enter plant specific means]

b. Hydrogen concentration - LESS

b. Consult plant engineering staff THAN 6.0% IN DRY AIR for additional recovery actions.

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 (4).

A faulted or ruptured SG should not be used in subsequent steps unless no intact SG is available.

9 Check Intact SG Levels:

a. Narrow range level - GREATER

o. Increase total feed flow to restore THAN (5)% [r6)% FOR ADVERSE narrow range level greater than d)%

CONTAINMENT]

[(6/% for adverse containment].

IF total feed flow less then (7) gpm, THEN go to Step 18. OBSERVE NOTE PRIOR TO STEP 18.

b. Control feed flow to maintain narrow range level between (5)%

[(6)% for adverse containment]

and 50%

4 of 10

M rm n.:-,m.

FR.C.1 RESPONSE TO INADEQUATE CORI COXING HP Rev.1 1 Sept.1983 ACTION / EXPECTED RESPONSE !

RESPONSE NOT OBTAINED STEP 10 Check RCS Vent Poths:

a. Power to PRZR PORV block

o. Restore power to block valves.

~

volves - AVAILABLE

b. PRZR PORVs - CLOSED

b. Manually close PRZR PORVs. E any volve can NOT be closed, THEN manually close its block valve.

c. Block volves - AT LEAST ONE

c. Open block valve unless it was OPEN closed to isolate an open PRZR PORV.

d. Other RCS vent paths - CLOSED

d. Close any open RCS vent path.

[ Enter plant specific listl NOTE Partial uncovering of SG tubes is acceptable in the following steps.

11 Depressurize Allintact SGs To (3) PSIG:

a. Dump steam to condenser at

a. Dump steam at maximum rate maximum rate using SG PORVs.

b. Check SG pressures - LESS

b. E SG pressure decreasing, THEN THAN /3) PSIG return to Step 9.E NOT, THEN go to Step 18. OBSERVE NOTE j

PRIOR TO STEP 18.

l

c. Check RCS hot leg temperatures -

c. H RCS hot leg temperatures AT LEAST TWO LESS THAN 400*F decreasing, THEN return to Step 9. E EQI, THEN go to Step 18. OBSERVE NOTE PRIOR TO STEP 18.

d. Stop SG depressurization 5 of 10

Messuuher TMos see. leeve/Deve FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev.1 1 Sept.1983 i

d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED 12 Check if SI Accumulators Should Be Isolated:

a. At least two RCS hot leg

a. Go to Step 18. OBSERVE NOTE temperatures - LESS THAN PRIOR TO STEP 18.

400*F i

b. Close all 51 accumalator

b. Vent any unisolated accumulator.

isolation valves 13 Stop All RCPs 14 Depressurize All Intact SGs To Atmospheric Pressure:

a. Dump steam to condenser at

a. Dump steam at maximum rate l

maximum rate using SG PORVs.

I 15 Verify SI Flov' Continue efforts to establish 51 flow.

Charging /51 pump flow indicators -

Try to establish any other high CHECK FOR FLOW pressure injection:

-OR-

[ Enter plant specific list].

High head 51 pump flow indicators -

IF core exit TCs less than 1200*F, CHECK FOR FLOW THEN return to Step 14. f NOT, THEN gt Step 18. OBSERVE NOTE PRIOR

-OR-TO STEP 18.

Low-head 51 pump flow indicators -

CHECK FOR FLOW 1

6 of 10

r pewauber:

TNies Rev. laswosDeses FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983, STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED H

16 Check Core Cooling:

a. Core exit TCs - LESS THAN 1200*F

a. Go to Step 18. OBSERVE NOTE PRIOR TO STEP 18.

b. At least two RCS hot leg

b. Return to Step 14.

temperatures - LESS THAN 350 F c.- RVLIS full range indication -

c. Return to Step 14 GREATER THAN (9) 17 Go To E 1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 12 NOTE Normai conditions are desired but not requiredfor starting the RCPs.

(

18 Check Core Exit TCs - LESS Start RCPs as necessary until core THAN 1200*F exit TCs less than 1200*F.

J.F, core exit TCs greater than 1200*F and all available RCPs running, THEN open all PRZR PORVs and block valves.

IF core exit TCs greater than 1200*F and all PRZR PORVs and block valves open, THEN open all other RCS vent paths to containment.

7 of 10

r

\\

s pewsiber:

71stos Rev. leave /Deve FR C.1 RESPONSE T0 INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE !

RESPONSE NOT OBTAINED 19 Try To Locally Depressurire All Use faulted or ruptured SG.

Intact SGs To Atmospheric Pressure:

Use PORY

-OR-

[ Enter plant specific means]

20 Check If SI Accumulators should Be isolated:

a. Low-head 51 pump flow

a. Return to Step 18.

Indicators - AT LEAST INTERMITTENT FLOW

b. Close all Si occumulator

b. Vent any unisolated isolation valves accumulator.

21 Check if RCPs $hould Be Stopped:

a. At least two RCS hot leg

a. Go to Step 22.

temperatures - LESS THAN 350*F

b. Stop all RCPs 22 Verify 51 Flow:

Continue efforts to establish Si Charging /SI pump flow indicators -

flow. Try to establish any other high CHECK FOR FLOW pressure injection:

t

-OR-

[ Enter plant specific list].

l l

High head 51 pump flow indicators -

Return to Step 18.

CHECX FOR FLOW

-OR-

Low-head Si pump flow indicators -

CHECK FOR FLOW l

8 of 10

- l

W W

Rev. leses/Deses FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP-Rev.1 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE '

{

RESPONSE NOT OBTAINED 23 Check Core Cooling:

Return to Step 18.

RVLIS full range indication -

GREATER THAN (9)

At least two RCS hot leg temperatures - LES THAN 350*F 24 Go To E 1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 12

- END -

l i

I 1

9 of 10

r js......o...:

j e

rm..

HP Rey,1 FR C.1 RESPONSI TO INADEQUATE CORE COOLING 1 Sept 1983l FOOTNOTES (1) Enter plant specific value corresponding to R WST switchover setpoint in plant specific units.

(2) Enter plant specific time.

(3) Enter plant specVic value which is 3-l/2 feet above the bottom of activefuelin core with :ero void fraction, plus uncertainties.

(4) Enter plant specific value correspondmg to CST low level switchover setpoint in plant specific units.

($) Enter plant specific value showing SG leveljust in the narrow range, including allowancesfor normal channel accuracy.

(6) Enter plant specific value showing SG leveljust in the narrow range, including allowancesfor normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 300*o.

(7) Enter the minimum safeguards AFWflow requirementfor heat removal, plus a;lowances for normal channel accuracy (typically one AfD AFWpump at SG design pressure).

(8) Enter plant specific value which is 200 psig, minus allowancesfor normal channel accuracy.

(9) Enter plant specific value which is above the top of activefuelin core with :ero voidfraction, plus uncertainties.

1 10 of 10

7 Attachment IV l

0290G:1/ GEL /4-85

i 1

%. L - ~.

NB IGT. l&44 dig /Mt E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sect.1983 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 25 Check if 51 Flow Should Be Reduced:

a. RCS subeccling based on core exit

a. 00 NOT STOP 51 PUMPS. Go to TCs - GREATER THAN tl.rj*F Step 27.

b. Secondary heat sink:

b. E neither cencition satisfied. THEN

Total feed flow to SGs - GREATER 00 NOT STOP 51 PUMPS. Go to T).'AN (61 GPM Step 27.

-CR-j

Narrow range level in at least one SG - GREATER THAN (B>?;

l

c. RCS pressure - STABLE OR

c. 00 NOT STOP 51 PUMPS. Go to INCREASING Step 27.

I

d. PRZR level - GREATER THAN /15/ 6

d. DO NOT STCP 51 PUMPS. Try to

?

sicbilize RCS pressure with ncrmal PRZR spray. Return to Step 25a.

26 Go To ES 1.1, Si TERMINATION, Step 1 l

9 of 13

ML20133K772

Contents

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ML20133K772

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Dear Mr. Denton:

Dear Mr. Denton:

SUMMARY

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