ML20203C355
| ML20203C355 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/11/1986 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Harold Denton, Tam P Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM 2NRC-6-037, 2NRC-6-37, TAC-62872, NUDOCS 8604210124 | |
| Download: ML20203C355 (76) | |
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'Af Duquesne Lidit 27,x3;;;g Nuclear Construction Divison Telecopy Itt$u$ HIP iS2 5 April 11, 1986 Mr. Ilarold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Mr. Peter Tam, Project Manager Division of PWR Licensing - A Office of Nuclear Reactor Regulations
SUBJECT:
Beaver Valley Power Station - Urit No. 2 Docket No. 50-412 Open Issue No. 3, Inadequate Core Cooling REFERENCE :
(a) Request for additional information, dated March 3, 1986.
(b) Westinghouse letter NS-NRC-86-3099, dated February 12, 1986 from E. P. Rahe to J. M. Taylor.
Gentlemen:
In response to your request for additional information (Reference a),
we are providing the following attachments which should address your concerns:
1.
Specific responses to your formal questions.
2.
A revised description of instrumentation for detection of inade-quate core cooling which supersedes our May 31, 1985 submittal.
The proposed responses to your request were discussed informally with your staff and we are confident this issue can now be resolved.
DUQUESNE LIGHT COMPANY By J. # Carey Vice President GLB/clk NR
/ COOL AR A Attachment cc:
Mr. R. Karsch - w/ attachment Mr. L. Prividy, Resident Inspector Mr. W. Troskoski, Resident Inspector
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DR ADOCK 050004,2 PDR
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION QUESTION NO.1:
Knowing that the CET accuracy may not be suitable for the calculation of the core subcooling margin, the applicant should justify the use of the CET signal as the only input to the RPU for the core subcool-ing margin.
Please address the status of the CET environmental qual ification.
RESPONSE :
As discussed in the attached description of instrumentation for detection of inadequate core cooling, the thermocouple inputs are used in the core subcooling margin calculation and the accuracy for this system is acceptable.
See Reference b) for additional infor-mation.
RTDs are now being employed in the RVLIS to obtain the necessary accuracy.
QUESTION N0. 2:
Describe any modifications or deviation from the Westinghouse Owner's Group Emergency Response Guidelines in the process of developing the Beaver Valley-2 Emergency Operating Procedures due to the uncertainty of the core exit thermoccuple readings.
RES PONSE:
No modifications or deviations from the WOG Emergency Response Guidelines are necessary.
Plant-specific setpoints will be used in the calculations.
QUESTION NO. 3:
Is the portion of the core exit thermocouple cabling inside the con-tainment environmentally qualified?
RESPONSE
Yes.
QUESTION NO. 4:
The description of the subcooling margin monitor signal processing electronics is not clear:
Are there two or three RPUs?
In parti-cular RPU-I is not shown to receive a signal from the reference junction boxes. Please clarify.
RESPONSE
There are four qualified (class IE) RPUs and one non-class IE RPU.
Inputs from the reference junction boxes go to two RPUs (III & IV).
Figure IV shows all of the RPUs. Refer also to the attached text.
QUESTION N0. 5:
In the RVLIS Figure V, it is not clear where the inputs to DPU A and OPU B are coming from. Please clarify.
RESPONSE
Figure V has been deleted.
Figure IV accurately addresses the 3
j QUESTION NO. 6:
In the description of the RVLIS monitoring system there is a High Volume Pressure Sensor; however, there is no indication of the pur-pose or use of this sensor signal.
Please describe.
RES PONSE:
The High Volume Pressure Sensors have been included on the revised Figure III (attached). A discussion of their function is included in the text of the RVLIS description.
Within the text they are called
" Sensor Bellows."
Their primary function is to provide isolation of the pressure sources.
QUESTION N0. 7:
What version of the WOG Emergency Response Guidelines has Beaver Valley adopted; i.e., Rev. 0,1, or 27 RES PONSE:
BVPS-2 is using Revision 1 of the WOG ERGS.
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Response to NUREG-0737. II.F.2
" Instrumentation for Detection of Inadequate Core Cooling" I.
The Inadequate Core Cooling (ICC) Monitoring System installed at Beaver Valley Unit 2 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 51 of the Beaver Valley Unit 2 chromel-alumel core exit thermocouples (26 on protection set III and 25 on protection set IV). A layout sketch of the system is shown in Figure I.
The train orientation of the thermocouples is shown in Table 1.
The core exit thermocouples are mounted at the top of the core support plate. They are then routed to four upper head conoseal penetrations. Af ter exiting the conoseal penetrations, the qualified thermocouple wires l
proceed through a swagelok and then to qualified connectors to i
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0398G:2/ GEL-BV2/ll85
facilitate disconnection during removal of the upper head. Upon exiting the reactor vessel cavity, the qualified cables are routed in a manner consistent with the reconsnendations 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) III and IV. The signals from both RPU's are routed to both Display Processing Units (DPU) for calculation of the compensated core exit themocouple 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 51 compensated thermocouple values, the information from both DPU's are transmitted to both seismically qualified flat panel Plant Safety Monitoring System (PSMS) displays.
DPU A and display A are powered by train A and DPU B and display B are powered by train B.
The cabling between the RPU's, DPU's and displays meet the recommendations of Regulatory Guide 1.75.
The system is qualified via individual component testing, which includes the cable inside containment.
B.
Core Subcoolina Marain Monitor The inputs to the core subcooling margin monitor include the following:
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Wide range RCS pressure (3 channels)
Core exit compensated thermocouple values (51 channels)
Reference junction box RTD values (6 channels)
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 (I, III, and IV). Also 26 uncompensated thermocouple channels and the corresponding 3 reference junction box RTD signals are input into RPU III (25 into RPU IV). 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. Use of the auctioneered high thermocouple quadrant average temperature in the calculation of subcooling margin is consistent with the utilization in the WOG Emergency Response Guidelines (ERG). The subcooling margin calculated values are routed to both displays (A and B).
The cable routing from sensor input to display meet the recommendations of Reg.
Guide 1.75.
The PSMS displays are the same display panels utilized in displaying the core exit thermocouple information.
0398G:4/ GEL-BV2/1185
C.
Reactor Vessel level Instrumentation System The Reactor Vessel Level Instrumentation System (RVLIS) consists of two redundant independent trains that monitor the water level in the reactor vessel.
The fluid diagram of one train of the Beaver Valley Unit 2 RVLIS system is shown in Figure III. The differential pressure (d/p) transmitters are located outside containment to minimize the large increase in measurement uncertainty (temperature, pressure, and radiation) associated with a change in the containment environment during an accident. The d/p transmitters are connected to the RCS penetration taps through a hydraulic system incorporating high volume sensors, capillary lines, hydraulic isolators, and manual isolation valves.
The pressure sensing lines connect to sealed capillary impulse lines which transmit the pressure measurements to the d/p transmitters located outside containment. The capillary impulse lines are sealed at the RCS end with a sensor bellows which serves as a hydraulic coupling for the pressure measurement. The impulse lines extend from the sensor bellows through the containment wall to hydraulic isolators which also functions as a bellows seal thus providing hydraulic coupling as well as isolation of the lines.
The capillary tubing extends from the hydraulic isolators to the d/p transmitters where instrument valves are provided for 0398G:5/ GEL-BV2/1185
trans:itt:r isolation. The capillaries are ir. stalled within instrument tubing channels to protect the capillaries from jet impingement and missiles where necessary.
The wide 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 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 approxinete indication of the relative void content of the circulating fluid. The inputs to the RVLIS system include the following:
1.
RCS hot leg wide range RTD's (2 channes1) 2.
Wide range RCS pressure (4 channels) 3.
Differential pressure (6 channels) 4.
Reference leg temperature values (12 channels) 5.
Reactor coolant pump status (4 channels)
The electrical block diagram associated with the RVLIS system is shown in Figure IV.
0398G:6/ GEL-BV2/1185
The RCS hot leg wide range RTD signals are input to RPU's III and IV. Also, one wide range RCS pressure channel is input into RPU I, III, and IV.
In addition, one of two sets of three differential pressure signals (wide range, narrow range, and dynamic head) are input into RPU III and IV, respectively. Also six reference leg compensating temperature inputs from each train of RVLIS are input into RPU III and IV.
Finally, to determine the appropriate RVLIS indication, the running status of each reactor coolant pump is input into the non-1E RPU N1.
Both trains of RVLIS readings are routed to both plasma displays.
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 the core subcooling margin and the core exit thermocouple information.
0398G:7/ GEL-BV2/1185
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 accident conditions. A description of the tests conducted and a discussion of the test results are presented in the following documents:
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.
0398G:8/ GEL-BV2/ll85
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.
5 0398G:9/ GEL-8V2/ll85
2 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.
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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 PSMS. The display package hierarchy, as summarized in Exhibit 1, includes the following:
1.
Top Level Plant Status Summary (Exhibit 2) 2.
Four Lower Level Graphic Displays a.
Core Temperature Map (Exhibit 3) b.
Pressure-Temperature Operating Limits (Exhibit 4) c.
Reactor Vessel Water Level (Exhibit 5) d.
Nuclear Power (Exhibit 6) 3.
Two Pages of Menu Display a.
Primary Data Trend, Secondary Data Trend and Containment Data Trend Menu (Exhibit 7) b.
Detailed Data Menu (Exhibit 15) 0398G:11/ GEL-BV2/1185
4.
Four Multi-Page Sets of Data a.
Five-Page Set of Primary Data Trends b.
Une-Page Set of Secondary Data Trends c.
One-Page Set of Containment Data Trends d.
Seven-Page Set of Detailed Data B.
The following exhibits provide a top down display of the core exit thermocouple information.
1.
a.
Exhibit 2 - maximum core exit thermocouple temperature.
b.
Exhibit 3 - 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 20 - spatially oriented core exit thermocouple map showing each thermocouple temperature.
d.
Exhibit 19 - Alpha numeric listing of core exit thermocouple location, tag designation and temperature reading per quadrant.
1 0398G:12/ GEL-BV2/1185
e.
Exhibit 9 - a two hour trend history of the three core exit thermocouple trisector maximum temperatures.
C.
The following exhibits provide a top down display of the core subcooling margin (based upon core exit thermocouples):
1.
a.
Exhibit 2 - core subcooling margin based upon core exit thermocouples.
b.
Exhibit 4 - RCS pressure - temperature plot exhibiting plant approach to saturation.
c.
Exhibit 16 - alpha numeric listing of core subcooling
- margin, d.
Exhibit 8 - 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 2 - displays appropriate RVLIS narrow and wide range and dynamic head readings depending upon RCP status.
0398G:13/ GEL-BV2/1185
b.
Exhibit 5 - mimic of analog meters indicating RVLIS narrow, wide and dynamic readings with respect to reactor vessel. Only displays appropriate ranges based upon RCP status.
c.
Exhibit 16 and C-10.3 - alpha numeric listing of appropriate ranges for both trains of RVLIS system.
d.
Exhibit 10 - a two hour trend history of all three RVLIS ranges. Also presents a trend of RCP status.
E.
Alarm Capability - The core exit thermocouple display pages are designed such that any numeric thermocouple readout greater than 1200"F will Le displayed ir, se video and flashed at a frequency of 1 hertz.
The core subcooling margin will indicate "SUBCOOL" 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 wl.en the auctioneered high quadrant thermocouple average temperature exceeds the coolant saturation temperature.
0398G:14/ GEL-BV2/ll85
F.
Backup Display - Since the Beaver Valley Unit 2 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.
The tima interval required for accessing the primary and backup display pages is within 5 seconds.
G.
Trend Capability - In addition to being displayed on the PSMS, the one train of RVLIS and one train of subcooling readings are recorded on the main control board. Also a channel of core exit thermocouple temperature is recorded on the main control board.
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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 2.
Core Subcooling Margin Monitoring Subsystem Document i
a.
Wide Range RCS Pressure ESE-2 b.
Core Exit Thermocouples See Item Above 0398G:16/ GEL-BV2/1185
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 g.
Microprocessors ESE-53 h.
Plasma Display ESE-61B 0398G:17/ GEL-BV2/Il85
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 Beaver Valley Unit 2 FSAR.
Included in the discussion is a justification for the number of channels selected and the diverse variable
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identified where necessary.
Presented in the Beaver Valley Unit 2 FSAR, Section 7.5, Table 7.5-1 is a detailed description of the i
characteristics associated with each ICC monitoring system, including range, number of channels, and qualification status.
C.
Power Supply - RPU I, DPU A and Display A are powered by inverter power bus I.
RPU II, DPU B and display B are powered by inverter power bus II.
RPU III is powered by inverter power bus III and l
l RPU IV is powered by inverter power bus IV.
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 2 display, and the detailed data display of Exhibits 3, 4, 5, 16, f
17,18,19, and 20. The recording capability of the ICCI channels is indicated in the Beaver Valley Unit 2 FSAR Table 7.5-1.
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E.
Quality Assur:Sce All hardware associated with the Beaver Valley Unit 2 PSMS ICCI monitoring systems meets the applicable portions of the quality assurance regulatory guides.
F.
Capability for Sensor Checks The Beaver Valley Unit 2 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, P00R, BAD and SUSPECT. The coerator 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 Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
H.
Channel Removal from Operation See Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
0398G:19/ GEL-BV2/1185
I.
Access to Setpoints Adjustments, Calibration and Test Points See Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
J.
Information Readout See Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
K.
System Repair See Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
L.
Derivation of System Inputs See Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
M.
Instrumentation Uti'lization To the extent practical, the Beaver Valley Unit 2 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 Beaver Valley Unit 2 FSAR, Section 7.5.2.3.
0398G:20/ GEL-BV2/ll85
1 VI.
Schedule The Beaver Valley Unit 2 ICCI monitoring system is to be installed and tested prior to fuel load.
Furthermore, the system will be calibrated prior to the plant achieving 5 percent power.
VII.
Beaver Valley Unit 2 is adopting the format and content of the Westinghouse Owners Group (WOG) Emergency Response Guidelines, Rev. 1 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 Beaver Valley Unit 2 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 Beaver Valley Unit 2 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 throughout the generic guidelines.
0398G:21/ GEL-BV2/ll85
VIII.
Duquesne Light should provide a discussion of the acceptability of the location of the PSMS displays based upon the results of the Beaver Valley Unit 2 Control Room Design Review.
0398G:22/ GEL-BV2/1185
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ATTACHMENT III 9
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0290G/ GEL /1-85
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Namator TWes l
see, seawe,Dese FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev.1 1 Sept.1983
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d STEP ACTION / EXPECTED RESPONSE }
{
RESPONSE NOT OBTAINED CA UTION
- Low-head SIpumps should not be run longer than a> without CCIV to the RHR heat exchangers.
1 Verify $1 Velve Alignment -
Manually align volves as necessary.
PROPER EMERGENCY ALIGNMENT 2
Verify 31 Flow in All Trains:
Start pumps and clign volves as
- Charging /SI pump flow indicators -
necessary. Try to establish any CHECX FOR FLOW other high pressure intection:
- High head Si pump flow indicators -
[ Enter plant specific list].
CHECX FOR FLOW Low-head 51 pump flow indicators -
CHECX FOR FLOW
/3 Check RCP Support Conditions -
Try to establish support conditions.
AVAILABLE
[ Enter plant specific list) l I
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FR C.1 RESPONSE TO INADEQUATE CORE C00 Lido HP Rev. I 1 Sept.1983 STEP}
ACTION / EXPECTED RESPONSE }
f 5dNSE NOT OBTAINED
}-
j
/4 Geck $1 Accumulator Isolation Velve 5tstva:
- a. Power to isolation volves -
)
AVAILABLE
- c. Restore pwer to isolation volves.
- b. Isolation volves - OPEN
- b. Open isolation volves unless closed offer occumulator discharge.
5 Geek Core fait TCs - LESS l
THAN 1200*F Go to Step 8.
6 Geck RVLl3 Full Range Indication:
- a. Indication - GREATER THAN (J)
- c. E increasing, THEN return to
- b. Return to guideline and step in Step 1. g NQT, THEN go to Step,7.
effect 7
Geck Core Izit TCs:
- a. Temperature - LESS THAN 700'F
- o. E decreasing THEN return to Step 1. g NOT, THEN go to Step 8.
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- b. Return to guideline and step in effect l
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Rev. laswomeos FR C.1 RESPONSE 70 INADEQUATE CORE COOLING HP.Rev. I 1 Sept.1983 STEP ACTION / EXPECTED RESPONSE }
RESPONSE NOT CBTAINED q
NOTE This guideline should be continued while obtainin hydrogen sample in Step 8.
3 Check Containment Hydrogen Concentration:
- a. Obtain a hydrogen concentration measurement:
[ Enter plant specific means]
- b. Hydrogen concentration - LESS THAN 6.056 IN DRY AIR
- b. Consult plant engineering staff for additional recovery actions.
Go to Step 9.
- c. Hydrogen concentration - LESS THAN 0.5Y. IN DRY AIR
- c. Turn on hydrogen recombiner system.
CA UTION
- Alternate water sourcesfor AFWpumps will be necessary if CST level decreases w less than ui.
9 Check intact SG Levels:
- a. Narrow range level - GREATER THAN (s)Y. [(6)ti FOR ADVERSE
- o. Increose total feed flow to restore CONTAINMENT]
narrow range level greater than (5/?;
[(6j'/. for adverse containment].
l JF total feed flow less then /7/ gpm, i
THEN go to Step 18. OBSERVE NOTE PRIOR TO STEP 18.
- b. Control feed flow to maintain narrow range level between (5)??
l
[(6)*6 for adverse containment]
and 50!?
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FR.C.1 RESPONSE TO INADEQUATE CORE COOLING.
HP Rev. I 1 Sept.1983 d STEP H ACTION /EXPECTEDRESPONSE l
[
RESPONSE NOT OBTAINED h
10 Check RC5 Vent Peths:
- a. Power to PRZR PORV block
- o. Restore power to block volves.
volves - AVAILABLE
- b. PRZR PORVs - CLOSED
- b. Manually close PRZR PORVs. g.
ony volve con NOT be closed, THEN manually close its block volve.
- c. Block volves - AT LEAST ONE
- c. Open block volve unless it was OPEN closed to isolate on open PRZR PORV.
- d. Other RCS vent paths - CLOSED
- d. Close any open RCS vent path.
[ Enter plant specific list)
NOTE Partial uncovering of SG tubes is acceptable in the following steps.
11 Depressurize AllIntact SGs To (B) P510:
- c. Dump steam to condenser at
- b. Check SG pressures - LESS
- c. Check RCS hot leg temperatures -
- c. iE RCS hot leg temperatures AT LEAST TWO LESS THAN 400*F cecreasing, THEN return to Step 9. E @_T, THEN go to Step 18. OBSERVE NOTE PRIOR TO STEP 18.
- d. Stop SG depressurization 5 of 10
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Eev. laswe ht FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED 12 Check if St Accumulators should Be isoleted:
- a. At least two RCS hot !eg
- o. Go to Step 18. OBSERVE NOTE temperaturer - LEM iHAN PRIOR TO STEP 18.
400*F
- b. Close all Si accumulator
- o. Dump steam to condenser at
15 Verify si Flow:
Continue efforts to establish Si flow.
Charging /SI 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 _
g core exit TCs less than 1200*F, CHECK FOR FLOW THEN return to Step 14.g NOT, THEN c
-OR-go to Step 18. CBSERVE NOTE PRIOR TO STEP 18.
~ Low head 51 pump flow indicators -
CHECK FOR FLOW 1
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n n.s-,m, FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 d STEP d ACTION / EXPECTED RESPONSE
- RESPONSE NOT OBTAINED m
16 Check Core Cooling:
- a. Core exit TCs - LESS THAN 1200*F
- o. 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
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NOTE Normal conditions are desired but not requiredfor starting the RCPs.
18 Check Cosa Exit TCs - LESS Start RCPs as necessary until core THAN 1200*F exit TCs less than 1200*F.
j,F. core exit TCs greater then 1200*F and all available RCPs running, THEN open all PRZR PORVs and block volves.
1F. core exit TCs greater than 1200*F and all PRZR PORVs and block volves open, THEN open all other RCS vent l
paths to containment, l
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FR C.1 RESPONSE T3 INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 d STEP m ACTION / EXPECTED RESPONSE l RESPONSE NOT 08TAINED h
19 Try To Locolly Depressurire All Use faulted or ruptured SG.
Intact SGs To Atmospheric Pressore:
- Use PORV
-OR-
- [ Enter plant specific means]
20 Check if SI Acetmulators should Be Isoleted:
~
- a. Low head Si pump flow
- o. Return to Step 18.
indicators - AT LEAST INTERMITTENT FLOW
- b. Oose all Si occumulator
- b. Vent any unisolated isolation volves occumulator.
21 Check if RCPs should Be Stopped:
- a. At lecst two RCS hot leg
- c. Go to Step 22.
temperatures - LESS THAN 350*F
- b. Stop all RCPs 22 Verify $1 Flow:
Continue efforts to establish 51 Charging /51 pump flow indicators -
flow. Try to establish any other high CHECK FOR FLOW pressure injection:
-OR-
[ Enter plant specific list].
- High head Si pump flow indicators -
Return to Step 18.
CHECK FOR FLOW
-OR-
- Low head 51 pump flow inclicators -
CHECK FOR FLOW 8 of to
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rue.
Soc. Seeue/Deser FR C.1 RESPONSE TO INADEQUATE CORE COOLING HP Rev. I 1 Sept.1983 STEP ACTION / EXPECTED RESPONSE }
RESPONSE NOT OBTAINED h-L 23 Check Core Cooling:
Return to Step 18.
- RVLIS full range indication -
GREATER THAN (9/
- END -
)
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FR C 1 RESPONSE TO INADEQUATE CORE COOLING l
1 Sept.1983;,
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FOOTNOTES (1)
Enter plant specific value corresponding to R WSTswitchover serpoint in plant specific units.
(2) Enter plant specipe time.
(3)
Enterplant speci)1c value which is 31/2 feet above the bottom of activefuelin core :vith :ero void fraction, plus ancertainties.
Enter plant specific value corresponding to CSTlow level switchover serpoint in plant specific units.
(4)
(3)
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 escred 30%.
(7)
Enter the minimum safegu.trds AFWflow requirementfor heat removal, p.'us allowances for normal channel accuracy (typically one AfD AFWpump at SG design pressure).
(8)
Enter plant specific value which is 200 psig, minus allowances for normal channel accuracy.
(9)
Enterplant specipe value which is above the top of activefuelin core with :ero voidfraction, plus uncertainties.
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Attachment IV 0398G:1/ GEL-BV2/1085
Menuhart TWes Rev.leeue 4hese E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED i-NOTE
- Steps 1 through 14 are IMMEDIA TE A CTION steps.
- Foldout page should be open.
1 Verify Reactur Trip:
Manually trip reactor. [F reactor will NOT Rod bottom lights - LIT trip, THEN go to FR 5.1, RESPONSE TO NUCLEAR POWER GENERATION / ATWS, Reactor trip cnd bypass breakers -
Sf8P I-OPEN Rod position indicators - AT ZERO Neutron flux - DECREASING 2
Verify Turbine Trip:
- a. All turbine stop valves - CLOSED
- a. Manually trip turbine.
(.
3 Verify Power To AC Emergency Busses:
- a. AC emergency busses - AT LEAST
- o. Try to restore power to at least one ONE ENERGlZED ac emergency bus. IF power can NOT be restored to at least one ac emergency b;s, THEN go to ECA-0.0, LOSS OF ALL AC POWER, Step 1.
- b. AC emergency busses - ALL
- b. Try to restore power to deenergized ENERGlZED ac emergency busses.
4 Check If Si is Actuated:
Check if 51 is required.1F. 5I is required,
[ Enter plant specific means]
THEN manually actuate. IF Si is NOT required, THEN go to E5 0.1, REACTOR TRIF RESPONSE, Step 1.
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E-0 REACTOR TRIP OR SAFETY INJECTION HP-Rev. I 1 Sept.1983 d STEP H_ ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED h
5 Verify FW lsoistion:
Manually close valves as necessary.
Flow control valves - CLOSED Flow control bypass volves - CLOSED FW isolation valves - CLO,5ED SG blowdown isolation volves -
=
CLOSED
+ SG sample isolation volves - CLOSED 6
Verify Containment isolation Phose A:
- a. Phase A - ACTUATED
- a. Manually actuate Phase A.
- b. Phase A valves - CLOSED
- b. Manually clcse volves.
7 Verify AFW Pumps Running:
- a. MD pumps - RUNNING
- a. Manually start pumps.
g
- b. Turbine-driven pump - RUNNING IF
- b. Manually open steam supply valves.
NECESSARY 8
Verify 51 Pumps Running:
Manually start pumps.
I Charging /51 pumps - RUNNING
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High-head 51 pumps - RUNNING Low-head 51 pumps - RUNNING 9
Verify CCW Pumps - RUNNING Manually start pumps.
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h IM*8 Sow, leanse/Deso:
E0 REACTOR TRIP OR SAFETY INJECTION HP Rev.1 1 Sept.1983 STEP ACTION / EXPECTED RESPONSE '
RESPONSE NOT OBTAINED
}-
10 Verify Service Water Pumps - RUNNING Manually start pumps.
11 Verify Containment Fen Coolers -
Manually start fan coolers in emergency RUNNING IN EMERGENCY MODE mode.
12 Verify Containment Ventilation isolation:
- a. Dampers - CLOSED
- a. Manually close dampers.
[ Appropriate steps for verification of other essential equipment as required by the specific plant design should be placed after Step 12.]
13 Check if Mein Steemlines should Be isolated:
- a. [ Enter plant specific means or
- a. Go to Step 14.
setpoints]
- b. Verify main steamline isolation and
- b. Manually close valves.
(
bypass valves - CLOSED 14 Verify Containment Spray Not Required:
- a. Containment pressure - HAS
- a. Perform the following:
REMAINED LESS THAN u) PSIG
- 1) Verify containment spray initiated.
y NOT, THEN manually initiate.
- 2) Verify containment isolation Phase B valves closed. E d.QI, T]EN manually close volves.
- 3) Stop all RCPs.
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E0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 STEP l l ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED 15 Verify $1 Flow:
- a. Charging /51 pump flow indicators -
- a. Manually start pumps and align CHECK FOR FLOW valves.
- b. Go to Step 16.
((3) PSIG FOR ADVERSE CONTAINMENT]
- c. High-head 5: pump flow indicators -
- c. Manually start pumps and align CHECK FOR FLud volves.
- d. Go to Step 16.
(r5) PSIG FOR ADVERSE CONTAINMENT]
- e. Low head 51 pump flow indicarors -
- e. Manually start pumps and align CHECK FOR FLOW valves.
16 Verify AFW Flow - GREATER THAN Manually start pumps and align valves as (6) GPM necessary. IF AFW flow greater than (6) gpm can NOT be established, THEN go to FR H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, Step 1.
17 Verify AFW Velve Alignment - PI'.0PER Manually align valves as necessary.
EMERGENCY ALIGNMENT l
18 Verify $1 Valve Alignment - PROPER Manually align valves as necessary.
l EMERGENCY ALIGNMENT l
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Nusuhart TMoe see, leanse/ Dees f.0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED
]
19 Check RC5 Average Temperature -
E temperature less than (71*F and STABLE AT OR TRENDING T0 (7)*F decreasing, THEN:
a) Stop dumping steam.
b) IF,cooldown continues, THEj control total feed flow. Maintain total feed flow greater than (6/ gpm until narrow range level greater than (s>.
((9;*'. for adverse containmer.t] in at least one SG.
c) { cooldown continues, THEN close main steamline isolation and bypass valves.
1 temperature greater than tr>*F and increasing, THEN:
Dump steam to condenser.
,(.
-OR-Dump steam using SG PORVs.
I s
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E0 REACTOR TRIP OR SAFETY INJECTION HP-Rev. I 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT 08TAINED 20 Check PRIR PORVs And Spray Valves:
- a. PORVs - CLOSED
- b. Normal PRZR spray volves - CLOSED
- b. IF PRZR pressure less than (111 psig, THEN manually close volves. E valves can NOT be closed, THEN stop RCP(s) supplying failed spray volve(s).
NOTE Sealinjection flow should be maintained to all RCPs.
21 Check if RCPs Should Be Stopped:
- a. 51 pumps - AT LEAST ONE RUNNING
- o. Go to Step 22.
- Charging /51
-OR-
- High-head 51
- b. RCP Trip Parameter - LESS THAN t/2>
- b. Go to Step 22.
[(13) FOR ADVERSE CONTAINMENT]
- c. Stop oil RCPs t
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stumher TWies eos. lesee/Deses I.0 REACTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT OSTAINED 22 Check if SGs Are Not Faulted:
- a. Check pressures in all SGs -
- o. Go to E-2, FAULTED STEAM
+ NO SG PRES 5URE DECREASING IN GENERATOR ISOLATION, Step 1.
AN UNCONTROLLED MANNER
Go to E 3, STEAM GENERATOR TUBE
- Condenser air ejector radiation -
RUPTURE, Step 1.
NORMAL
Go to E-1, LOSS OF REACTOR OR
- Containment radiction - NORMAL SECONDARY COOLANT, Step 1.
- Containment pressure - NORMAL
(
- Containment recirculation sump level
- NORMAL I
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E.0 REACTOR TRIP OR SAFETY INJECTION HPJev.1 1 Sept.1983 d STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT 08TAINED 25 Check if Si Flow Should Be Reduced:
- o. RCS subcooling bosed on core exit
- a. DO NOT STOP Si PUMPS. Go to TCs - GREATER THAN (14>'F Step 27.
- b. Secondary heat sink:
- b. IE neither condition satisfied. THEN
- Total feed flow to SGs - GREATER DO NOT STOP St PUMPS. Go to THAN (6) GPM Step 27.
-OR-Narrow range level in at least one a
SG - GREATER THAN (31%
- c. DO NOT STOP Sl PbMPS. Go to INCREASING Step 27.
- d. PRZR lovel - GREATER THAN //5/*6
- d. DO NOT STOP St PUMPS. Try to stabilize RCS pressure with normal
,e PRZR spray. Return to Step 250.
k 26 Go To IS.I.1, Si TIRMINATION, Step 1 I
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phemmer Thies Bow, leeuesDooos E0 REArTOR TRIP OR SAFETY INJECTION HP Rev. I 1 Sept.1983 STEP H ACTION / EXPECTED RESPONSE l RESPONSE NOT O8TAINED F
27 laitiate Monitoring Of Critical Safety Fenetion Status Trees CA UTION Alternate water sources for AFWpumps will be necessary if CST level decreases to less than asi.
28 Check SG Levels:
- a. Narrow range level - GREATER THAN
- a. Maintain total feed flow greater than (31%
(6) gpm until narrow range level greater than (3/% in at lecst one SG.
- b. Control feed flow to maintain narrow
- b. IF narrow range leve! in any SG range level between (31% and 50%
continues to increase in an uncontrolled manner, THEN go to E 3, STEAM GENERATOR TUBE RUPTURE, k.
Step 1.
29 Check Secondary Radiation - NORMAL Go to E 3, STEAM GENERATOR TUBE (Enter plant specific means]
RUPTURE, Step 1.
30 Check Auxiliary Building Radiation -
Evaluate 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.
31 Check PRT Conditions - NORMAL Evaluate cause of abnormal conditions.
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E0 REACTOR TRIP OR SAFETY INJECTION HP Rev.1 1 Sept.1983 d STEP }--
ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED CA UTION If offsite power is lost after SI reset, manual action may be required to restart safeguards equipment.
32 Roset$1 33 Reset Containment isolation Phose A And Phase 8 34 Establish lastrument Air To Containment Start one air compressor and establish instrument air to containment.
CA UTION RCS pressure should be monitored. If RCS pressure decreases to less than u> psig the low-head 51 pumps must be manually restarted to supply water to the RCS.
35 Check if Low. Head 51 Pumps should Be Stopped:
- a. Check RCS pressure:
- 1) Pressure - GREATER THAN
- 2) Pressure - STABLE OR INCREASING
- 2) Go to Step 36.
- b. Stop low-head 51 pumps and place in standby
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E0 REACTOR TRIP OR SAFETY INJECTION HP Rev.1 1 Sept.1983
_m STEP d ACTION / EXPECTED RESPONSE l RESPONSE NOT OSTAINED 36 Check if Diesel Generators should Be Stopped:
- a. Verify oc emergency busses -
- c. Try to restore offsite power to oc ENERGlZED BY OFFSITE POWER emergency busses. E offsite power con NOT be restored, THEN load the following equipment on oc emergency busses:
(Enter plant specific list).
- b. Stop any unloaded diesel generator and place in stondby 37 Return To Step 19
- END -
b i
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E0 REACTOR TRIP OR SAFETY INJECTION HP-Rev.I 1 Sept.1983
(
l FOOTNOTES 1
l (1) Enter plant specific containment pressure serpoint for spray actuation.
l (2) Enter plant specific valuefor the shutoff head prenure of the high-head 51 pumps, plus allowances for 1
1 normal channel accuracy.
1 (3) Enter plant specific value for the shutoff head pressure of the high-head SI pumps, plus allowances for normal channel accuracy and post accident transmitter errors, not to exceed 2000 psig.
(4) Enter plant specific valuefor the shutoff head pressure of the low-head Sipumps, plus allowances for normal channel accuracy.
($) Enter plant specific valuefor the shutoff head pressure of the low-head Sipumps, plus allowances for normal channel accuracy and post accident transmitter errors.
(6) Enter the minimum safeguards AFWflow requirement for heat removal, plus allowances for normal channel accuracy (typically one SfD AFWpump capacity at SG design pressure).
(7) Enter plant specific no-load temperature.
(8) Enter plant specific value showing SG leveljust in the narrow range, including allowances for normal channel accuracy.
(9) Enter plant specific value showing SG leveljust in the narrow range, including allowancesfor normal channel accuracy, post accident transmuter errors, and reference leg process errors, not to exceed 30%.
(10) Enter PRZR POR Vpreuure setpoint.
I I
(11) Enter PRZR Spray pressure setpoint.
(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 serpoint, including allowances for normal channel accuracy and post accident transmitter errors. Refer to document RCP TRIP!RESTARTin Generic issues sec-tion of Esecutive Volume.
(14) Enter sum of temperature and pressure measurement system errors, mcluding allowancesfor normal channel accuracies, translated onto temperature using saturation tables.
!!!) Enter plant specific value showmg PRZR leveljust in range, mcluding allowances for normal channel
- accuracy, 116) Enter plant specific value corresponding to CSTlow level switchover serpoint in plant specific unus.
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