ML19296D300
| ML19296D300 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 02/22/1980 |
| From: | Peoples D COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8003030353 | |
| Download: ML19296D300 (16) | |
Text
Commonwealth Edison one First National Plaza. Chicago. Ittinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 February 22, 1980 t.
Dr. H. R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Station Units 1 and 2 Documentation of compliance with Short Ters Requirements of the Lessons Learned Task Force NRC Docket Nos. 50-295/304
Dear Dr. Denton:
During the January 21, 1980, visit to Zion Station by Mr. John Olshinski and others of your staff, a request was made for additional documentation of our compliance with the Category A items of NUREG-0578.
The enclosed responses are provided as a result of that request.
One (1) signed original and thirty-nine (39) copies of this transmittal are provided for your use.
Very truly yours, D. L. Peoples Director of Nuclear Licensing Enclosures 80 0p30 3C
2.1.1 As discussed in our January 1, 1980, response the power source for the pressurizer heaters, in the event of loss of normal off-site power, is from the emergency diesel generators.
The pressurizer heater loads were included in the normal and emergency (including LOCA) design diesel generator loading and will not be automatically shed from the emergency diesel bus upon occurrence of a safety injection actuation signal.
2.1.3.a With regard to item 3 of the October 30, 1979, clarifi-cation letter, the backup to the single channel direct indication of PORV position is an alarm on high temperature in the PORV discharge line.
In the emergency procedure that would be followed for a small break LOCA, the operator is directed to check this alarm in addition to checking the stem-mounted position indicators.
In addition to the high temperature alarm on the PORV relief line, an alarm actuated by the PORV stem-mounted position indication will be installed as soon as possible, but no later than April 1, 1980.
With regard to acoustical cross-talk as related to an open PORV, the acoustical monitoring system installed on the pressurizer safety valve discharge lines may alarm for a stuck open PORV.
This presents no problems to the operator since he is directed by procedures to check the alarm on high temperature in the PORV relief line and the direct indication of the PORV position to determine if the alarm is associated with a stuck open PORV.
If a stuck open PORV is indicated, the operator would be directed to isolate the PORV and subsequent operator actions would be identical no matter whether it was a stuck open safety valve or PORV which initiated the ala rm.
2.1.3.b At Zion Station subcooling indications are displayed on the main control board in the form of core-exit temperature, saturation temperature based on RCS pressure, saturation pressure based on core temperature, and degrees subcooling.
The indications are derived by the unit's process computer using core-exit thermocouple temperatures and reactor coolant system pressure as input.
Operating personnel have been instructed not to make operational decisions based on this single plant parameter when confirmatory indications are available.
The digital process computer based system is a testable and highly reliable indicator of core subcooling.
Last year the average process computer availability at Zion was 99%.
The subcooling indications are re-computed every 32 seconds and are continuously displayed on the meters on the main control board.
These calculations are performed routinely and are not bypassed, even if the alarm sequence typewriter is overloaded.
Thus, no significant computational delay is expected during the course of an accident.
Core-exit temperature is derived by the unit's process computer from the average of the 10 highest indicating thermocouples out of an array of 65 thermocouples which measure temperature just above the reactor core.
Reactor coolant system pressure is obtained from the average of 4 safety grade narrow range pressure channels between 1750 and 2500 psig and from a single non-safety grade wide range pressure channel for all other pressures between 0 and 3000 psig.
In the event one of the narrow range pressure channels fails or deviates substantially from the reading of the other 3 channels, the computer would automatically reject the suspect reading. A second wide range pressure channel will be input to the process computer by January 1, 1981, and both of the non-safety grade wide range pressure channels will be modified to the extent practicable to improve their environmental qualifications.
In the interim, the operator will be cautioned by procedure not to place reliance on information derived from the wide range pressure channel, if containment conditions exceed the qualified range of the instrument.
The process computer converts the reactor coolant system pressure to saturation temperature and then subtracts from it the rea' tor coolant system temperature as obtained from the average of the 10 highest reading incore thermocouples.
This gives the degrees of subcooling which is displayed on the main control board indicator.
The process computer also converts the core-exit temperature to a saturation pressure which is displayed on the main control board indicator.
The inputs to the process computer from the four narrow range pressure channels, the wide range pressure channel, and the incore thermocouples are sent through appropriate signal isolation to prevent any interference with the operation of reactor protection or engineered safeguards-systems.
Other reliable instrumentation and procedures are available to the operator such that subcooling can be determined manually in the unlikely event that the process computer is unavailable during a transient.
All core exit thermocouples can be read manually on a permanently-mounted meter in the control room.
Two hard-wired, analogue indications of reactor pressure are displayed and recorded on the control board.
Similar hard-wired meters give continuous, recorded analogue indication of temperatures in the hot and cold legs of each of the four reactor coolant loops.
These '.emperatures are derived from RTDs mounted directly in the main loop piping.
Otl' narrow range RTDs are mounted in manifolds in small diameter bypase lines on each leg.
The analogue output of these RTDs is continuously displayed on the control board as delta T across each loop.
The proposed arrangement of continuous display of computer calculated saturation conditions, backed by manual determination of subcooling from a variety of instruments hard-wired to the control room appears to satisfy NUREG-0578 requirements.
Attached is a completed questionnaire which was transmitted by the NRC letter dated October 30, 1979.
INFORMATION REQUIRED ON THE SUBC00 LING METER Display Information Displayed Tsat, Teore, Psat, (T-Tsat, Tsat, Press, etc.)
Subcooling F Display Type (Analog, Digital, CRT)
Analog Continuous or on Demand Continuous Single or Redundant Display Single Location of Display Control Room Alarms (include setpoints)
None Overall uncertainty ( F, PSI)
Normal containment conditions Tsat: 1F Teore: 10 F Psat: 170 psi Subcooling: 11 F Post-accident contain-ment conditions Tsat: 14 F Teore: 16 F Psat: 280 psi Subcooling 30 F Range of Display Tsat: 500-700 F Teore: 500-700 F Psat: 0-3000 psi Subcooling: 0-100 F Qualifications (seismic, liigh quality environmental, IEEE-323) commercially available Calculator Type (process computer, dedicated Process computer digital, or analog calc.)
If process computer is used, 99%
specify availability (% of time)
Single or redundant calculators Single Selection Logic (highest T.,
10 highest indicating lowest press)
T/Cs read from an array of 65 T/Cs Qualifications (seismic, None environmental, IEEE-323)
Calculation Technique (Steam Steam Tables Tables, Functional Fit, ranges)
Input Temperature (RTD's or T/Cs)
T/Cs Temperature (number of sensors and 65 above core locations)
Range of temperature sensors 0-2250 F on process computer 0-700 F on control board indicator ( to be increased to 0-2300 F as soon as possible)
Uncertainty of temperature sensors Normal containment conditions: 3 F at 670 F*
Post-accident contain-ment conditions: 9F at 670 F*
Qualifications (seismic, Seismic and some environmental, IEEE-323) environmental 0
- 670 7 corresponds to saturation temperature at me.ximum primary system pressure (i.e., safety valves relief pressure)
~5-Pressure (specify instrument used)
Narrow range RCS pressure and wide range RCS pressure Pressure (number of sensors and Four narrow range locations)
(pressurizer) and one wide range (RHR hot leg letdown line)
Range of Pressure sensors 1700-2500 psig narrow range; 0-3000 psig wide range Uncerts.inty of pressure sensors Normal containment con-ditions:
narrow range: 8 psi wide range: 30 psi Post-accident contain-ment conditions:
narrow range: 211 psi wide range: unknown Qualifications (seismic, Narrow range: IEEE-323-environmental, IEEE-323) 71 and seismic Wide range: none Backup Capability Availability of Temp & Press
- 1) All core exit T/Cs read out in control room.
- 2) Two analog indications of reactor press, on CNTL board.
- 3) Analog indication of each loop hot and cold leg temp.
- 4) Delta T across each loop.
Availability of Steam Tables, etc.
Yes Training of operators Complete Procedures Implemented 2.1.4 I.
The following containment penetrations were identified as needing further documentation to show compliance with the NUREG position and general design criteria for containment isolation systems:
1.
RCP Seal Water Return This penetration was designated as non-essential in our response of January 1, 1980.
Automatic isolation is provided by closure of one MOV on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 2.
FSAR Section 6.6.2 states that Class 2 penetrations include outgoing lines not missile protected, or which can otherwise communicate with the containment atmosphere following an accident.
In general, the minimum isolation capability required for Class 2 penetrations is two automatic isolation valves in series, with automatic seal water injection. The RCP seal water return line is provided with special design features in order to achieve the required degree of redundancy. Two independent isolation signals, one from each of the redundant safeguards logic trains A and B, are sent to the motor operated isolation valve.
The valve itself is a double disc type of gate valve.
As described in Note 5 of FSAR 6.6.2, this type of valve when sealed by water injection, provides two barriers against leakage. Thus, the isolation capability provided for the RCP seal water return line meets the requirement for this class of penetration.
2.
Cooling Water Return from RCP This penetration was designated as essential in CECO's response of January 1, 1980, and therefore automatic isolation is not required by the NUREG-0578 item 2.1.4 position. However, automatic isolation is provided by closure of one MOV on a Phase B isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 2.
The required degree of redundancy is provided by the double disc motor operated gate valve, supplied with isolation valve seal water injection.
Thus, the isolation capability provided for the cooling water return from the RCP meets the requirement for this class of penetration.
3.
RCP Seal Water Supply This penetration was designated as essential in our response of January 1,1980, and therefore automatic isolation is not required by the NUREG-0578 item 2.1.4 position.
Isolation capability is provided by two manual valves in series, with provision for manual isolation valve seal water injection.
FSAR Table 6.6.5.1 classifies this penetration as Class 3A.
FSAR Section 6.6.2 states that Class 3A penetrations include incoming lines not missile protected, or which can otherwise communicate with the containment atmosphere following an accident. Acceptable isolation arrangements for Class 3A penetrations include two manual isolation valves in series, with seal water injection.
The isolation capability provided for the RCP seal water supply meets this requirement.
4.
Containment Air Particulate and Gas Monitor Outlet This penetration was designated as non-essential in our response of January 1,1980. As documented in FSAR Table 6.6.5.1, this incoming line has a missile-protected check valve inside the containment which acts as the primary barrier.
The system is closed outside the containment to provide a second barrier. A manual isolation valve outside the containment can also be used as a second barrier.
This arrangement was found acceptable at the time Zion's operating license was issued.
It has been reviewed in light of the TMI incident and still appears to be the best for Zion's sampling configuration.
This is a common return for sampling lines and must be open to accomplish any sampling of the containment atmosphere.
If power-operated isolation valves had been included on this line, a single failure to open would defeat all the air sampling systems.
The current arrangement is therefore the most reliable.
In fact, since containment air sampling is essential to accident recovery, this penetration shsuld be reclassified as essential.
II. The following containment penetrations were identified as needing further documentation to show compliance with the NUREG position:
l '.
Cooling Water Return from Excess Letdown Heat Exchanger This penetration was designated as non-essential in our response of January 1, 1980. Automatic isolation is provided by closure of one A0V on a Phase A isolation s ignal.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
FSAR Section 6.6.2 states that Class 4 penetrations include incoming and outgoing lines which are connected to closed systems inside containment and are protected from missiles throughout their length.
The minimum isolation capability required for Class 4 penetrations is one manual isolation valve located outside the containment.
The isolation capability provided for the cooling water return from the excess letdown heat exchanger exceeds this requirement.
2.
Cooling Water Supply to Excess Letdown Heat Exchanger This penetration was designated as non-essential in our response of January 1, 1980.
Isolation capability is provided by a check valve inside containment backed up by a manual valve located outside of containrent.
The system is closed inside containment and pressure is automatically maintained at a value greater than peak containment accident pressure by pumps driven from essential AC power.
In addition the operator may isolate the primary side of the excess let-down heat exchanger using valves remotely controlled from the main control room. Note 4 of FSAR Section 6.6.2 states that a check valve qualifies as an automatic trip valve in certain incoming lines not requiring seal water injection.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
The isolation capability provided for the cooling water supply to
'the excess letdown heat exchanger exceeds the requirement for this class of penetration.
3.
Fire Protection to Containment This penetration was designated as non-essential in our response of January 1, 1980.
Automatic isolation is provided by closure of one A0V on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
The isolation capability provided for the fire protection supply exceeds the requirement for this class of penetration.
4.
Steam Generator Blowdown Sample (4 penetrations)
These penetrations were designated as non-essential in our response of January 1, 1980.
Automatic isolation is provided by closure of one A0V on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies these penetrations as Class 4.
The isolation capability provided for the steam generator blowdown sample lines exceeds the requirement for this class of penetration.
5.
Nitrogen Manifold to Pressurizer Relief Tank This penetration was designated as non-essential in our response of January 1,1980. Automatic isolation is provided by closure of one A0V on a Phase A isolation signal, and a manual valve is provided in series with this A0V.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
The isolation Capability provided for this nitrogen supply to the pressurizer relief tank exceeds the requirement for this class of penetration.
6.
RC Loop Fill Header This penetration was designated as non-essential in our response of January 1, 1980.
Isolation capability is provided by two administratively controlled, normally locked closed manual valves, with isolation valve seal water injection supplied on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 3A.
FSAR Section 6.6.2 states that Class 3A penetrations include incoming lines not missile protected, or which can otherwise communicate with the containment atmosphere following an accident.
Acceptable isolation arrangements for Class 3A penetrations include two manual isolation valves in series, with seal water injection.
The isolation capability provided for the loop fill header meets this requirement.
When operations require opening the valves during power operation, a dedicated individual, in communication with the control room, shall be available to close the valves in an emergency or upon termination of the operation.
7.
Purification Pump to Refueling Cavity This penetration was inadvertently omitted from our response of January 1, 1980.
It is designated as non-essential.
Isolation capability is provided by two administratively controlled, normr11y locked closed manual valves, with isolation valve seal water injection supplied on a Phase A isolation signal.
This penetration is classified as Class 3A.
The isolation capaoility provided meets the requirement for this class of penetration. When operations require opening the valves during power operation, a dedicated individual, in communication with the control room, shall be available to close the valves in an emergency or upon termination of the operation.
8.
Refueling Cavity Drain This penetration is more properly identified as the purification pump line from the refueling cavity.
This penetration was designated as non-essential in CECO's response of January 1,1980.
Isolation capability is provided by two administratively controlled, normally locked closed manual valves, with isolation valve seal water injection
_to-supplied on a Phase A isolation signal.
This penetration is classified as Class 3A.
The isolation capability provided meets the requirement for this class of penetration.
When operations require opening the valves during power operation, a dedicated individual, in communication with the control room, shall be available to close the valves in an emergency or upon termination of the operation.
9.
Containment Air Sampling Inlet (4 penetrations)
These penetrations were designated as non-essential in our response of January 1, 1980.
Isolation capability is provided by two administratively controlled, normally locked closed manual valves.
FSAR Table 6.6.5.1 classifies these penetrations as Class 5.
FSAR Section 6.6.2 states that Class 5 penetrations include lines which can be opened to containment atmosphere, but which are normally closed during reactor operation. Two manual isolation valves in series are required for this class of penetrations.
The isolation capability provided for the containnent air sampling lines meets this requirement. When operations require opening the valves during power operation, a dedicated individual, in communication with the control room, shall be available to close the valves in an emergency or upon termination of the operation.
- 10. Nitrogen Manifold to RCDT This penetration was designated as non-essential in our response of January 1, 1980.
Isolation sapability is provided by a check valve and the automatic closure of one A0V on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
The isolation capability provided for the nitrogen supply to the RCDT exceeds the requirement for this class of penetration.
- 11. Nitrogen Supply to Accumulators This penetration was designated as non-essential in our
~
response of January 1, 1980.
Isolation capability is provided by a check valve and the automatic closure of one A0V on a Phase A isolation signal.
FSAR Table 6.6.5.1 classifies this penetration as Class 4.
The isolation capability provided for the nitrogen supply to the accumulators exceeds the requ'.rement for this class of penetration.
2.1.6.a The Leak Reduction and Control Program at Zion Station covers the following systems: Waste Gas System, H 2 Recombiner, Containment Air Sampling System, ECCS Piping (SI, Charging, RHR),
and Reactor Coolant Sample Piping.
These include all the systems listed on page A-26 of NUREG-0578, and they comprise all systems outside containment thst would or could contain highly radioactive fluids during a serious transient or accident.
The containment air sample system leakage of 97.5 SCFH is higher than desired.
There are many reasons for this leakage, including sample pump shaft seals.
Reduction of the leakage requires equipment replacement which will be completed by January 1, 1981, as commited in the response to NUREG-0578, Item 2.1.8.a.
Improved Post-Accident Sampling Capability.
A leak detection program for liquid radwaste systems at Zion Station has been in effect since mid-1978.
The program entails the daily use of a data sheet.
Twenty-one pump run time meter readings and twelve tank levels are recorded on the data sheet.
A flow chart of the radwaste water flow is used to identify the daily total flow through the radwaste system.
From this data a daily water balance is calculated. The daily data collected is posted on a monthly flow calculation chart.
This helps in spotting trends and abnormalities in total flow and individual pump flows.
Any abnormal flows or increases are investigated and proper action taken as needed.
2.1.6.b One vital area at Zion Station was inadvertently omitted from our January 1, 1980, response.
Startup of the hydrogen recombiner requires access to the containment purge exhaust plenum, prior to system lineup, for attachment of a blankoff plate in the suction piping.
The shielding study which evaluated doses in that plenum, determine that access is acceptable provided modifications are done to the system ss described in Item 2.1.5 of our January 1, 1980, response.
2.1.8.a Interim procedures implemented to meet the requirements of Section 2.1.8.a include :
(a) ZCP 500 " Post Accident Sampling and Analysis", (b) ZCP 23A " Boron Analysis Procedure Under Post Accident Conditions" and (c) ZCP 123A " Hydrogen Analysis of Gas Samples."
These procedures describe methods (a) to sample the post accident reactor coolant and to analyze for isotopic activities and boron, and (b) to sample the containment atmosphere for isotopic activities and hydrogen.
ZCP500:
The reactor coolant is sampled in the primary sample room.
A second sample is collected after a 5 minute purge of the sample lines to the VCT.
The sample is transported to the laboratory for dilution and analysis.
The boron analysis is done either by (a) titration method using mannetol (ZCP
- 23) or (b) a colorimetric method using a curcumin reagent
. (ZCP 23A).
Portable shielding and at least 2 personnel will be used to keep radiation exposures within NRC limits.
The containment atmcaphere will be sampled from the R11/R12 containment radiation monitor.
A 2-mf11111ter sample will be collected in a vacutainer tube using a syringe (normally used for blood collection). One (1) ml of the sample will be analyzed for isotopic activities using the GeLi gamma ray spectrometry.
The rest of the sample will be used for hydrogen analysis using the Fisher-Hamilton Gas Partitioner.
ZCP23A Boron analysis is done by forming a color complex with a curcumin rangent and by measuring the intensity of the color developed with a spectrophotometer.
ZCP123A Hydrogen analysis is done by using a Fisher-Hamilton Gas Partitioner with helium gas as the carrier. A 1% hydrogen gas standard is used. The hydrogen concentration is determined by comparing the hydrogen peak areas of the standard and the sample.
The training of the Rad / Chem technicians in the above procedures has been completed.
2.1.8.b Procedure RP 1740-1 " Monitoring High Activity Gaseous Releases During Accident Conditions" describes the real time monitoring system for estimating noble gas releases from the auxiliary building and potential releases from the containment and condenser air ejector pathways.
In addition, the procedure provides a methodology for assessing radioiodine and particulate releases. A high range noble gas effluent monitor already exists for gas decay tank releases.
The Emergency Operating Procedure for the SG tube rupture accident addresses the collaction and analysis of steam samples to help quantify releases from the atmospheric relief dumps.
Noble gas releases are quantified by making exposure rate measurements 6 inches from appropriate sampling lines for each release path.
A response curie is provided for converting the exposure rate (mR/hr) into concentration (uCi/cc).
This curve is based on a line dose calculation which takes into account the radionuclide distribution with time af ter the accident.
Subsequently, the release rate is then calculated based on the exhaus t rate.
Personnel monitoring the release are in direct communication with the control room.
This monitoring can be done every 15 minutes.
Survey meters which would be used to make these measurements are calibrated once per quarter using a CS-137 source.
The instruments have a relatively flat energy response for the range of energies being measured.
Using this method the monitoring sensitivity 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the accident is approximately 20 uCi/cc.
. For auxiliary building noble gas releases the individual will monitor the release from the 617' elevctic, in the auxiliary building. A lead collimator is used to reducu the background and improve the monitoring sensitivity. The sample will be drawn through tygon tubing from the appropriate tack monitors which are located on the 642' elevation in the auxiliary building.
This sampling scheme is necessary because the auxiliary building stack monitors are located near the HEPA and charcoal filter banks.
During an accident these filters could result in a very high exposure field near the effluent monitors.
During an accident the tygon tube connections can be made in under 30 seconds per monitor (one for each unit).
This methodology will result in the lowest occupational exposure.
For the containment and condenser air ejector release paths this monitoring is done at the appropriate rad monitor.
The containment effluent monitors are located in the unit 1 and 2 purge rooms (617' elevation auxiliary building).
The air ejector monitors are located on the 592' elevation in the turbine building.
Exposure rates will not prohibit making these measurements from either location.
The auxiliary building stack monitors are supplied from normal AC power while the containment effluent monitors are supplied from essential buses.
An alternate back-up AC supply powered from the unaf fected unit will be available, if necessary, to allow sampling of the auxiliary building effluent.
For radioiodines and particulates, instructions are provided for obtaining cartridge and filter samples from the appropriate monitors under accident conditions. The quantification of releases will be made using GeLi isotopic analysis of the samples.
The detector cave will be purged with clean air to reduce the xenon background.
The iodine cartridges are blown out for at least 10 minutes to reduce the level of entrapp:d noble gases.
In addition, silver zeolite cartridges are on order and when available their use will significantly reduce the interference from noble gases.
The sampling locations are the same as those describsd above for the noble gas monitoring. Only the auxiliary build tng ef fluent samples present an occupational exposure problem.
Esposure permitting these samples will be obtained from the monitor, otherwise they will be obtained from the 617' sampling location.
The latter sampling location would affect the particulate activity measurement but should have little impact on the iodine measurement.
Calculational methods for determining release rates will be made using existing station chemistry procedures.
This information will be relayed immediately to the control room and Technical Support Center.
The GeLi/MCA equipment is calibrated annually and checked at least quarterly.
2.1.8.c Two portable Eberline SAM-1 radioiodine measurement systems are on order.
Expected delivery is late February, 1980.
Procedures for the operation and calibration of this equipment will be written at that time.
Procedure RP 1740-3 describes the methods by which airborne radioiodine samples are safely obtained under accident conditions.
The procedure requires that the iodine cartridges be blown out for at least 10 minutes to reduce the level of entrapped noble gases.
In addition, the use of silver zeolite cartridges (curreatly on order) is required to further reduce the interference from high noble gas activity.
Until the arrival of the SAM-2 (single channel analyzer) equipment, these iodine samples will be analyzed using the GeLi MCA in the counting room.
When the SAM-2 is received, RP 1740-3 will be revised to incorporate its use.
2.1.9.e The RCS Vessel Vent System was sized to assure that there would not be any LOCA concern due to either the planned use or in-advertent operation of the system.
The Zion FSAR page 14.3.1-1 at:Les the changing pump flow is sufficient to maintain system pressure above the Safety Injection trip point through a 1/2" I.D.
opening. A 1/2" pipe will be used for the Vent System to prevent operation of the system from adversely affecting the capability of the charging system to maintain RCS pressure.
Inadvertent operation will be prevented by placing the vent valves under administrative control with key-lock control switches.
The manual valve will be normally in the open position.
Normal system venting during start-up will be through either the solenoid valves, or through an additional manual valve added to the vent header by a T section which is not in line with the vent solenoid valves.
It would not be possible to continue normal venting operations during start-up with the manual isolation valve closed. The vent system is designed to become an integral part of the vessel head. The manual valve will be closed only to allow for isolation during any maintenance operations on the system.
Two RTDs or thermocouples, depending on availability, will be placed in the discharge pipe downstream from the solenoid vent valves.
These will be independently powered and connect to an alarm window on the control board to indicate system actuation.
The analysis to determine the alarm set point has not yet been c omple ted.
Only safety grade materials are to be used for both the ciectrical and mechanical components of the system.
All materials will meet or exceed the present requirements for Zion Station.
Electrical connections will be subject to IEEE 323-74 requirements.
An analysis of the vent rate is being conducted and will be available the week of March 3, 1980.
.