ML19309E603
| ML19309E603 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 04/15/1980 |
| From: | Broehl D PORTLAND GENERAL ELECTRIC CO. |
| To: | Harold Denton, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-30680, TAC-30681, TAC-30682, TAC-30683, TAC-30684, TAC-30685, TAC-30686, TAC-30687, TAC-30688, TAC-30689, TAC-30690, TAC-30691, TAC-30692, NUDOCS 8004220695 | |
| Download: ML19309E603 (26) | |
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.<a April 15, 1980 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Sir:
Attached please find supplemental information regarding NUREG-0578 Short-Term Lessons learned Actions, which was discussed and requested by members of the h3C Staff for official documentation during the telephone conference on March 24, 1980 and the meeting on March 27, 1980 between PGE and the NRC.
The attached information supplements previous PGE submittals to the NRC on this subject: submittals made on October 17, November 20, December 7, December 20, 1979, and January 2, 1980.
Sincerely, W
a o c:
Mr. A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors Mr. Lynn Frank, Director State of Oregon D3 7 Department of Energy
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i Section 2.1.1 - Emergeniy Power Supply Requirements for the Pressurizer Heaters. Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs The motive power for the PORVs is air which is provided from the station instrument air supply system through solenoid valves. The PORV solenoid valves are powered from redundant Class 1E station batteries in order to be independent of the availability of offsite power. Each PORV (total of j
2 PORVs at Trojan) is equipped with Seismic Category I air supply accumu-lators which are sized to support approximately 30 PORV operations in the l
event the normal nonseismic air supply is lost or unavailable. The air supply accumulators are located inside the Containment to be automatically t
repressurized via check valves from the instrument air supply system without any manual action.
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Section 2.1.3.a - Direct Indication of Safety Valve Position -
The acoustic monitoring system installed at the pressurizer safety valves is composed of three separate channels; each channel is dedicated to one pressurizer safety valve. Each channel consists of 1) an accelerometer, mounted approximately 6 in. downstream from the valve, 2) a charge con-verter, located near the accelerometer, which translates the charge into a voltage signal, and 3) an instrument rack module, located in the con-trol room, which translates the voltage signal into a discrete indication of valve position via light-emitting diodes indicating relative valve flow. The relative valve flow is indicated based on the phenomenon of acoustical accelerations set up by flow through the discharge piping near the valve.
The alarm setpoint was originally set at 25 percent of flow level based on the manufacturer's (Technology for Energy Corporation) recommendation.
This alarm level was chosen to be high enough to prevent false alarms due to background noise and low enough to ensure an alarm before the valve is fully open.
Recent experience with the accelerometers conducted by the manufacturer at operating plants has indicated that 1) the background level around the safety valves is extremely low, 2) interference between each sensor is negligible due to minimal crosstalk with adjacent valve.
The manu-facturer has informed us that previous concerns of the false activation of alarm due to low setpoint is not necessary.
Based upon Section 15.3 of the Trojan FSAR, the maximum leak flow for which normal makeup can maintain pressurizer level at operating pressure is calculated to be 17.5 lb/sec., which is predicted to occur for a 3/8-in. dianater small-break LOCA. The pressurizer safety valve dis-charge line at Trojan is 6 in. in diameter, which could discharge approxi-mately 117 lb/sec. of saturated steam per each valve.
Considering the discharge capacity of the valve for a break of 3/8-in. diameter, it was conservatively concluded that the alarm setpoint could be set as low as 9 percent without initiating frequent false alarms but still being less than the normal make-up capability. This alarm set point of 9 percent will be implemented prior to start up of Cycle 3 operation in 1980.
As an indirect indication of pressurizer relief and safety valves, the current Trojan plant Off-Normal Instruction (ONI-36) describes the pro-cedure to be followed for detecting leaks or actuation of the valves by checking pressurizer eclief tank high temperature, pressure and water level, and high temperature in the pressurizer relief and safety valve discharge.
KM/4mg6.lA19 Section 2.1.3.b - Instrumentation for Detection of Inadequate Core Cooling Subcooling Margin Meters (SMMs)
The temperature signals are obtained from RTDs in each hot leg and cold leg of each coolant loop and from 16 incore thermocouples (T/Cs). The pressure signals are obtained from two wide-range pressure sensors in the RHR suction leg of reactor coolant loop 4 Both pressure signals are fed to both SMMs in order to provide redundant pressure inputs.
The pressure sensors are Class IE systems and satisfies the cafety-grade design requirements in accordance with the Trojan FSAR commitments. Electric power to the SHMs is supplied from battery-backed vital instrument buses and is channelized for each train. The indication is provided at the electronics drawers on control room panels C09A and C09B and at main control board C12.
The panel racks on both C09 and C12 are qualified for Seismic Category 1 design requirement.
Following is additional information requested in the NRC letter of October 30, 1979 on the SMMs.
Display Uncertainty (*F, psi)
Digital Meter 13*F based on RTDs, 14*F based on T/Cs Analog Meter 15'F for both RTDs and T/Cs Qualifications (seismic, Regulatory Guide 1.97 (Revision 2) environmental, IEEE-323)
In process of qualifying to 1EEE 344-1975 Calculator Qualifications (seismic, Regula'cory Guide 1.97 (Revision 2) environmental, IEEE-323)
In process of qualifying to IEEE 344-1975 Input Uncertainty of temperature 123.6*F for RTD; 18.2*F for T/C sensors (*F at 1)
Qualifications (seismic, Review in progress; will be environmental, IEEE-32J) provided by lby 31, 1980 Uncertainty of pressure 160.9 psi wide range pressure sensor sensors (psi at 1)
Qualifications (seismic, Review in progress; will be environmental, IEEE-323) provided by :Sy 31, 1980
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-s Section 2.1.4 - Containment Isolation The Essential Systems are defined to be those fluid system lines pene-trating the Containment Building which must remain functioning in order to bring the Plant into a safe shutdown condition following a Design Basis Accident. Based on this definition, the fluid lines included in the essential systems are the Type IV fluid lines for the ESF systems, the four Containment pressure sensing lines, the hydrogen vent lines, feedwater lines and Reactor Coolant Pump seal injection line. Trojan plant FSAR Section 6.2.4 and FSAR Table 6.2-1 defines the Type IV fluid lines for the ESF systems: Residual Heat Removal, Component Cooling Water, Centrifugal Changing Pump, Safety Injection and Containment Spray Systems. With the exception of the hydrogen vent line, the essential system valves will not be isolated by Containment isolation signal. A summary of the Containment isolation barriers provided for all fluid systems penetrating the Containment is delineated in Trojan plant FSAR Table 6.2.1 of which a copy was transmitted to the NRC in PCE response dated October 17, 1979.
Due to the modification implemented for the control circuits of 23 Con-tainment isolation valves, the current Trojan design of cor.crol systems for automatic isolation valves does not reopen the isolation valves when the Containment isolation signal is reset.
Reopening of each Containment isolation valve will require deliberate operator action on the individual valve control switch.
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Section 2.1.5.a - Dedicated Penetrations for External Recombiners or Post Accident Purge System Hydrogen concentration of the Containment atmosphere is measured by the redundant, Seismic Category I post-accident Hydrogen Analysis System (FSAR Section 6.2.5 and FSAR Figure 6.2-48).
The maximum measurement range of the hydrogen analyzer is 10 percent hydrogen in air.
The Hydrogen Analysis System is designed to operate through the Containment pressure range of 0-60 psig. The indication of the hydrogen concentra-tion is displayed in a remote indicator which is mounted on the Seismic Category I panel C41 in the control room.
With regard to operation of hydrogen recombiners, the Trojan Emergency Instruction (EI-1) specifies the operator to initiate operation "if the RCS pressure was below saturation for some period of time and/or RCS samples show cladding damage and/or a buildup of hydrogen gas".
See Section 2.1.8.a of this Attachment for additional information on the operation of the Containment Hydrogen Analysis System.
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Section 2.1.6.a - Integrity of Systems Outside Containment Likely to Contain Radioactivity
)
The supplemental leak reduction program had been developed at Trojan to reduce leakage from systems outside Containment that would or could contain highly radioactive fluids during a serious transient or accident.
The program consists of periodic leak tests on specified systems to verify system integrity and to identify leakage.
Where possible, the testing will be performed by inspecting potential leakage areas such as flanges, pump seals and valve packing while the system is maintained at operating pressure. For systems containing gases, the test program will include the use of helium leak tracer.
The supplemental leak reduction program will be implemented at refueling cycle interval.
Weld and piping integrity will continue to be tested in accordance with the Inservice Inspection Program.
Leakage identified during the test will be properly recorded and a schedule for repair of tie source of excess leakage will be established to ensure its timely completion consistent with system operation and the magnitude of the leak. The acceptance criteria to be applied for leakage will be the design leak rate for individual components within the l
system.
Systems and/or portions of systems that would contain highly radioactive fluids during a serious transient or accident condition were reviewed for necessity of supplemental leak detection. The criteria used in determin-ing which systems to be included in the supplemental program were:
1.
System or a portion.of the system located outside of the Containment.
2.
System likely to contain highly radioactive fluids in the event of a serious ~ transient or accident.
3.
System or a portion of the system required to function
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during or following an accident or severe transient in order to bring the Plant to a safe shutdown condition.
By applying these criteria, the following systems were included in the leak reduction program:
Chemical and Volume Control System 1-including letdown, makeup,
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seal injection, CVCS deminerali:ers and holdup tanks.
Safety Injection System - flow-path from RRR to C"CS charging.
pump suction via the safety injection system.
Residual Heat Removal System - all of the system outside of Containment., --
9 Containment Spray System - all portions of the system required to recirculate water from the sump.
Radioactive Gaseous Waste System - all portions of the system associated with processing the gas from the CVC3 holdup tanks.
Reactor Coolant Sampling System - Portions of the sampling system required for post-accident sampling.
Other Plant systems were excluded from the program since they did not meet the three criteria listed above. Examples of those systems excluded are:
Boron Recovery System - not required for safe shutdown of the
]
Plant.
Liquid Radioactive Waste System - not required for safe shutdown of the Plant.
(This system was excluded by the NRC in the NRC letter of clarifications dated October 30, 1979.)
Chemical and Volume Control System - not required for safe shutdown of the Plant.
(That portion of the system isolated and not included in the above list.)
Component Cooling Water System - not containing radioactivity during normal and accident condition. Two pressure boundary must fail before leakage to the environment.
Hydrogen Recombiners - system within Containment.
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Section 2.1.6.b - Design Review of Plant Shielding and Environmental Qualification A summary of design review of plant shielding methodology and preliminary conclusions is presented in PGE response to NRC dated January 2, 1980.
The following information supplements that earlier response:
1.
Post-accident dose rates at emergency power supplies and the Containment isolation reset switches were evaluated and found to be acceptable for personnel access.
Emergency power supplies are located in the Control Building and Turbine Building, well away from areas of significant dose rate. Containment isolation can be reset from the control room where the dose rates have been evaluated to be within the requirements of CDC 19 of 10 CFR 50.
2.
Systems assumed to operate post-accident for the purposes of shielding and environmental qualification of equipment are essentially the same as the systems presented in Section 2.1.6.a.
As noted in our January 2, 1980 submit-tal, two operating modes were considered:
post-LOCA and the intact Reactor Coolant System (RCS) modes.
Systems assumed to operate during the post-LOCA mode included the Residual Heat Removal System, the Safety Injection System and the Containment Spray System, all in the recirculation mode.
Systems or portions of systems assumed to operate with internal high-activity fluids during the intact RCS.
mode include:
a.
Chemical and Volume Control System (CVCS) letdown to the holdup tanks. Prior to letdown to the holdup tanks, the CVCS demineralizers are assumed to be valved out of service.
Reactor coolant pump seal injection will be provided from a nonradioactive source if required.
Hydrogen and noncondensible degassing of the RCS, if required, will be accom-plished by using the Reactor Pressure Vessel Vent System and by pressurizer power-operated relief valve discharge to the pressurizer relief tank.
b.
Although originally evaluated, Trojan will not rely on letdown to the volume control tank for degassing the RCS.
As mentioned in a. above, de;assing the large volumes and noncondensibles such as hydrogen will be acccmplished through the Reactor Pressure Vessel Vent System. Total degas-sing of noble gases is not considered necessary during the initial phases of the accident.
e c.
The entire Radioactive Caseous Waste System was assumed to contain high levels of noble gas radioactivity.
3.
Field run piping was included in the design review of plant shielding presented in our January 2, 1980 response.
At Trojan, field run piping includes piping 2 in. and smaller which is Safety Class 3 or 4.
Examples of field run piping that have been evaluated include:
the reactor coolant sample lines through the sample room and back to the volume control tank; residual heat removal sample lines to the sample room; the holdup tank gaseous vent line to the waste gas surge tank; piping from the waste gas surge tank through the waste gas compressors to the waste gas decay tanks.
4.
A preliminary design review of the environmental qualifi-cation of equipment at Trojan has been conducted and areas for additional evaluation have been identified. This review was conducted to ensure that the equipment will perform its intended safety function during post-accident radiation exposure conditions.
The electrical equipment qualification data generated during ongoing review of IE Bulletin 79-01B was utilized in this review.
This review was based on the methodology and post-accident operation conditions developed in the January 2, 1980 response and operating conditions noted above.
The integrated dose to the equipment was determined using the following assumptions:
a.
The source terms in each system were the same used for the shielding evaluation with the exception of the Waste Gas System. The more realistic gaseous source term resulting from the filling of three CVCS holdup tanks was assumed instead of considering that all waste gas from the RCS was degassed through the-volume control tank.
b.
The dose was calculated for each component type' (e.g., valve, pump) which was judged to have the highest dose in each room.
c.
The integrated dose included both beta and gamma radiation if the material of interest was in contact with the process fluid. Otherwise, only gamma radiation was considered.
d.
As listed below, the total dose was determined for the =aximum time period each system was anticipated to operate. _
e Residual Heat. Removal System:
1 yr.
Containment Spray and Safety Injection Systems:
1 yr.
CVCS letdown to the holdup tanks:
I week Waste Gas System actual degassing time from the holdup tanks:
1-1/2 days Waste gas decay tanks:
1 yr.
The following are the preliminary results of this review:
a.
The dose to equipment located outside rooms containing the radioactive systems is within the limit of 4 x 10) rads in FSAR Section 3.11.
b.
The dose to valve motor operators, pump motors and pump shaft seals within the RHR, Containment Spray and Safety Injection Systems was less than the equipment qualifica-tion provided by the manufacturer.
c.
Additional evaluations are being performed for valve packing and "0" ringn, solenoid valves and air operator diaphrams, diaphrams in diaphram sected valves, gaskets and level switches.
Modifications, if required, will be completed prior to January 1, 1981. TO'.;/JLT/ng/4sa631 -
Section 2.1.7.a - Auxiliary 7eedwater Flow Indication to Steam Generators As discussed in PGE responses to D. G. Eisenhut on October 17, 1979 and to H. Denton on January 2,1980, redundancy in determining indirect auxil-iary feedwater flow exists at the Trojan plant by using the safety-grade channels of flow measurement and indication of the water levels in each steam generator. Additionally, safety grade ITT Barton flow indicating switches are provided locally on each of the eight auxiliary feedwater lines. Attached Table I provides additional information on the power supply matrix for auxiliary feedwater flow indicators as well as steam generator water level indicators.
l It should be noted that although the power supply to the four auxiliary flow indicators is currently provided from vital a-c bus Y13 (refer to power supply matrix in Table I), a design change will be made for sepa-rate power suppplies; flow indicators (FI3043 A2 and C2) for steam generators A and C will be powered from vital a-c bus Y13 and the remain-ing flow indicators (FI3043 B2 and D2) from vital a-c bus Y24. This modification will be completed before startup for Cycle 3 operation in 1980.
Subsequent to the PGE response of October 17, 1979, supplemental informa-tion was transmitted to the NRC (A. Schwencer) on December 14, 1979 for auxiliary flow indication to steam generators. As indicated in this PGE submittal, flow transmitters and indicating instruments are calibrated every refueling outage to maintain an accuracy of 11/2 percent of span for the transmitters and 12 percent of span for the indicators. Addi-tionally, the routine channel check of the flow indication system is conducted every 31 days.
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e Sheet 1 of 2 TABLE I Power Supply Matrix for Auxiliary Feedwater Flow and Steam Generator Level Indications Steam Steam Steam Steam Title Generator A Generator B Generator C Generator D Auxiliary Feedwater Flow at Panel C05 Indicator FI 3043A2 FI 3043B2 FI 3043C2 FI 3043D2 Power Supply Y13 Y13 Y13 Y13 Steam Generator Level (Wide Range) at Panel C05 Recorde r LR 501 LR 501 LR 503 LR 503 Power Supply Y24 Y24 Y24 Y24 i
Steam Generator Level (Narrow Range) at Panel C14 Indicator LI 517 LI 527 LI 537 LI 547 Power Supply Y24 Y24 Y24 Y24 i
Stean Generator Level (Narrow Range) at Panel C14 Indicator LI 518 LI 528 LI 538 LI.548 Power Supply Y13 Y13 Y13 Y13 Steam Generator Level
(" arrow Range) at Panel.C14 Indicator LI 519 LI 529 LI 539.
LI 549 Power Supply Y22.
Y11 Y11 Y22-
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Sheet 2 of 2 TABLE I Steam Steam Steam Steam Title Generator A Generator B Generator C Generator D Auxiliary Feedwater Flow Local Indica-tion at Main Steam Support Structure at Elevation 59 ft. 6 in.
Indicator FIS 3004A1 FIS 3004B1 FIS 3004Cl FIS 3004D1 Power Supply NA*
NA*
NA*
NA*
Indicator FIS 3004A2 FIS 3004B2 FIS 3004C2 FIS 3004D2 I
Power Supply NA*
NA*
NA*
NA*
- These flow indicators are differential pressure gages and thus do not require electrical power to operate the indication.
NOTE: Y13 and Y11 are vital a-c buses powered by train "A"
battery.
Y22 and Y24 are vital a-c buses powered by train "B"
battery.
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3 2.1.8.a - Improved Post Accident Sampling Capability - Interim Procedures Ir*erim sampling procedures which will allow analysis of reactor coolant anu Containment atmosphere during periods of severe core damage have been developed at the Trojan Nuclear Power Plant. The procedures are
,.s-d e-the reactor coolant source term specified in short-term lessons learned item 2.1.6.b.
Refined interim sampling procedures which assure addi-tional radiation controls are currently being generated and will be in place prior to Cycle 3 startup.
The refined interim sampling procedures are being developed by first cal-culating expected dose rates in the vicinity of the Containment atmospheric monitoring sample racks and reactor coolant sample sink. Then, evaluat-ing operator action required to take samples and estimating stay times and distance from sources such as 3/8-in. reactor coolant sample lines and grab sample septums. Once the stay times and operator actions were established, total exposure was evaluated and modifications to sampling equipment and analysis procedures developed to minimize operator exposure.
As a result of modification to procedures and equipment, reactor coolant and Containment atmospheric sampling can be accomplished at Trojan with exposure to chemistry and radiation protection personnel of less than 3 rem total body and 18 rem to the hands.
Summary of sampling method-ology, chemistry analysis and equipment modifications is presented below with emphasis on radiation controls.
Containment Atmospheric Sampling and Analysis Containment atmosphere is monitored for hydrogen, noble gas and iodine.
Hydrogen analysis is accomplished by redundant hydrogen analyzers located in an accessible area in the Auxiliary Building. The analyzers are con-tinuously in a hot standby condition, ready to accept Containment atmos-pheric samples. Hydrogen analysis is accomplished by opening the sample line isolation valves from the main control room and dispensing an Auxiliary Building operator to depress a pump start button on the hydro-gen analysis panel C285A and 3.
The hydrogen analyzer reads up to 10 percent hydrogen remotely in the main control room.
An attachment fitting has been installed at the hydrogen analyzer cabinet C2853 to allow the installation of a grab sample septum and silver eolite cartridge holder to allow sampling noble gas and iodine.
In the event of an accident, the sample rack is attached to the cabinet piping prior to initiating recirculation. Two noble gas samples will be taken by syringe frca a septum, placed into a lead pig for transfer to the counting room and hot lab.
The silver zeolite cartridge has a quick disconnect which allows removal into a shielded pig for transport to the counting room for_ analysis.
The silver zeolite cartridge will be counted by a gamma spectrometer in the hot lab.
If the gamma spectrometer is off scale, lead bricks and
_ distance will be used to reduce the count rate.
Finally, if this fails, an lon chamber will be used to measure dose rate and a dose rate versus Ci/cc curve will be used to determine activity concentrations.
While circulating Con'tainment atmosphere through the silver zeolite cartridge, 1
continuous radiation monitoring is performed to assure that radiation levels in the cartridge do not exceed 5 rem /hr. The silver zeolite cartridge is handled by use of tongs to further minimize operator exposure. External dose rates at the face of the hydrogen cabinet are calculated to be less than 10 rem /hr., I hr. after the accident and less than 2.6 rem /hr. I day after the accident. The grab sample rack is constructed of 3/8-in. tubing with.065-in. wall thickness, thus allowing for small source volumes.
Noble gas analysis will be performed by taking a syringe sample and counting it in the GeLi detector.
If the detector saturates, distance and shielding will be used. As a backup, an ion chamber will be used and a dose rate versus Ci/cc conversion factor will be used to determine activity concentration. Estimated contact dose rates on the sample syringe with 30-min. decay on the Containment atmospheric sample are calculated to be less than 15 rem /hr. The syringe must be handled only during sample withdrawal.
Reactor Coolant Sampling and Analysis Reactor coolant sampling procedures are in place for taking a liquid sample from the sample sink for analysis in the hot lab and for taking a hydrogen gaseous sample from the " hydrogen sample rig" located at the sample sink.
l The reactor coolant liquid and reactor coolant dissolved gas sampling both require access to the sample sink in the sample room. Estimated general area dose rates in the sample room while the reactor coolant sample lines are on recirculation are as follows:
General area dose 54 tsm/hr.
18 rem /hr.
rate at entry L'
to sample room 30 min.
24 hr.
Area dose rate 155 ren/hr.
52 rem /hr.
2 ft. from reactor 0
0 coolant sample 30 min.
24 hr.
spigot Area dose rate at 1077 rem /hr.
364 rem /hr.
2 ft. from 0
0 hydrogen sample 30 min.
24 hr.
rig The chemists taking samples will be alerted to potential high radiation levels in the sample room by a permanently-installed area radiation monitor located just inside the sample room.
Valves requiring operation at the sample sink are being fitted with reach rod capability which will allow operation.at a minimum of 2 ft. distance.
This capability will be installed prior to Cycle 3 startup.
The operator will never be c1cser than 2 ft. from the piping and valves containing 9
reactor coolant. Preliminary time-motion studies have confirmed that the necessary samples can be obtained within 1 day by incurring less than 3 rem whole body,and less than 18 rem to hands to any one chemist.
The estimated dose rates 2 ft. from a 5 ml reactor coolant sample are 18 rem /hr. with 30-min. decay and 6 rem /hr. with 24-hr. decay. Reactor coolant samples will be handled with a minimum 2-ft. long tongs. The 5 ml sample will be placed into a shielded container using 2-ft. long tongs and the sample will be transferred to a shielded fume hood in the hot lab. Hot lab gamma analysis and boren analysis will be performed on reactor coolant samples. Gamma analysis will be performed by placing i ut of reactor coolant into a 1-liter flask, causing a 1,000,000 to one dilution of reactor coolant.
The 1 ut sample will be obtained using an 12-in long micro pipette. The diluted reactor coolant sample will be counted in a gamma spectrometer located in the counting room.
Calculated background dose rates in the counting room due to direct radiation is negligible. Ventilation for the counting room maintains a positive pressure relative to the hot lab and sample room.
If the counting room background becomes too high, the counting equipment can be moved to a low background area.
Reactor coolant boron analysis will be set up and performed in a fume hood ana lead bricks will be available to provide additional shielding.
The chemistry procedure is similar to a normal boron analysis procedure used at Trojan; that is, a pipette is used to place.25 to 1 mi of reactor coolant into a 50 mi glass beaker set up on a magnetic stirring apparatus. A titration assembly using extended leads and an extended titrant delivery tube will be used in conjunction with a pH meter to perform boron analysis. This assembly will allow analysis at a distance of 5 ft.
Reactor coolant dissolved gas analysis will be performed by taking a 2 cc gaseous sample from the sample septun located within the existing hydrogen sample kit at the sample sink. The sample will be taken by a syringe and placed into a shielded sample container. The expected contact dose rate from the syringe is less than 100 rem /hr. at 24-hr. decay. The gaseous sample will be injected into a gas chromatograph for hydrogen analysis.
The gas chromatograph vents through Tygon tubing to a ventilation hood.
The sampling and analyses described above will be outlined in refined interim chemistry procedures. Chemistry personnel trained to perform these procedures can be onsite and complete the sampling and analysis within 24 hr.
Radiation work permits will be filled out and reviewed by radiation protection management personnel prior to entry into any high radiation area during post-accident sampling.
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Section 2.1.8.b Increased Range of Radiation !!onitors NRC POSITION 2 Procedures for Quantifying High-Level Accidental Radioactivity Releases Noble Cases The attached Table summarizes the sampling procedures which were in place prior to January 1, 1980.
Subsequent changes in the procedures which are planned also are described in the Table.
Temporary area radiation monitors (AR!!) will replace the grab sample method of determining the noble gas release rates for the Containment Ventilation Exhaust and Condenser Air Ejector Exhaust Systems.
This change is being made to further minimize personnel exposure during sampling and to provide additional assurance that noble gas release rates can be determined at least every 15 min. if necessary.
The ion chamber detector for the AK!! will be located on the PER!! sample line and shielded to reduce background radiation levels. The AR!! readout will be located away from major radiation sources to ensure post-accident accessibility.
The temporary ARM is capable of measuring effluent concentrations of up to 105 pCi/cc Xe-133.
The ion chamber will measure gamma ray energies as low as 80 kev and will be calibrated quarterly using a cesium 137 source in accordance with existing plant procedures.
The arf 1 is a-c powered with a battery backup.
The measured dose rate will be converted into uCi/cc Xe-133.
Readings will be transmitted to the control room via the inplant telephone system..
Crab sampling is retained for the Steam Valves and Power Operated Relief Valves.
Direct measurement with an ARM is not considered to be feasible because of the low Xe-133 gamma ray energy and the shielding provided by the heavy piping materials in the steam system. Personnel exposures for this sampling should not be excessive because of the steam dilution of the radioactivity.
The Plant is ' currently shu't down for refueling.
In the etent of an acci-dent during the outage, radiation exposures associated with grab sampling should not be excessive.
However, the temporary ARMS will be installed and procedures for cheir use issued prior to the start of. Cycle 3.
Applicable personnel will be trained in the use of the ARMS by the same date.
Chemistry and Radiation Protection Technicians training in the effluent sampling and analysis procedures is in progress and will be completed prior to startup of Cycle 3.
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NOBLE CAS SAMPLING PROCEDURES Release Point
_ !!ethod i
Containment Ventilation Exhaust System (l)
Grab Sample (2)
Auxiliary and Fuel Building Ventilation Exhaust Grab Sample (4)
System (3)
Condenser Air Ejector Exhaust System Grab Sample (2,5)
Wast < Gas Decay Tank Not Required (6)
Steam Safety Valves and Power-Operated Relief Grab Sample f
Valves I
(1) Includes Hydrogen Purge Exhaust System.
(2) Prior to start to Cycle 3 a temporary ARM will replace the grab sample method.
l (3) CVCS letdown, Emergency Core Cooling Systems and radwaste t
facilities are located in the Auxiliary and Fuel Building.
(4) A high-range detector for this PERM (300 uCi/cc Xe-133)
]
will be installed prior to start of Cycle 3.
Grab sampling will not be required after the high range detector is installed.
(5) A high-range detector for this PERM (300 uCi/cc Xe-133) will be installed prior to the start of. Cycle 3.
(6) Existing PERM has upper range of 105 uC1/cc Xe-133.
Discharge from Waste Gas Decay Tanks is to Auxiliary and Fuel Building Ventilation Exhaust System.
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Section 2.1.8.c - Improved Iodine Instrumentation In-Plant airborne radioiodine samples will be collected on a charcoal or silver zeolite cartridge using a standard portable low volume air sampler.
The iodine on the cartridge will be measured using an Eberline SAM-Il port-able two-channel analyzer which is calibrated to the 365 key Iodine-131 gamma ray.
Since the low volume air samplers are portable, samples can be collected throughout the Plant and in all vital areas such as the control room, Technical Support Center and Operational Support Center.
Since the Eberline SAM-II analyzer is portable, it can be located in low background areas which eliminates the requirement for special shielding.
Silver zeolite cartridges will be used if noble gas interference is suspected.
Use of silver zeolite cartridges precludes the need to flush the cart-ridges with a clean gas prior to analysis.
Procedures are issued for both obtaining and analyzing in-Plant airborne t
radioiodine samples.
Chemistry and Radiation Protection Technicians have been trained in these procedures.
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Section 2.1.9 - Reactor Coolant System Venting In accordance with the telephone conversation on March 24, 1980, the following is the supplemental information which responds item-by-item to Section C "PWR Vent Design Considerations" specified in the NRC letter of clarifications (Pages 48-50 in Enclosure 1) dated October 30, 1979.
1.
The locations for PWR vents are as follows:
a.
Each PWR licensee should provide the capacity to vent the reactor vessel head.
PGE Response The Remote Head Vent System to be provided at Trojan is an extension of the existing local reactor vessel head vent piping and will provide the capacity to vent the reactor vessel head, b.
The reactor vessel head vent should be capable of venting noncondensible gas from the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and the cold legs (through head jets and other leakage paths).
l PGE Response l
l The Trojan Head Vent System will vent noncon-densible gases from hot-leg pipes to the top of
]
the outlet nozzles and will vent the cold legs j
through the head jets in the core barrel support plate.
c.
Venting of the pressurizer is required to assure its availability for system pressure and volume control.
PGE Response The Trojan pressurizer can be vented to the pres-surizer relief tank through the existing Power-Operated Relief Valves (PORVs).
2.
The size of the reactor coolant vents is not a critical issue.
The desired venting capability can be achieved with vents in a fairly large range of sizes. The criteria for sizing a vent can be developed in several ways. One
- approach, which we consider reasonable, is to specify a volume of noncondensable gas to be vented and a venting time; i.e., a vent capable of venting a gas volume of one half the RCS in I hr.
Other criteria and engineer-
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ing approaches shouldfbe considered if desired. _
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PCE Response The Trojan Head Vent System is designed to be capable of venting one half the RCS volume in 1 hr.
t 3.
Where practical, the RCS vents should be kept smaller than the size corresponding to the definition of a LOCA (10 CFR 50, Appendix A).
This will minimize the chal-lenges to the ECCS since the inadvertent opening of a vent smaller than the LOCA definition would not require ECCS actuation, although it may result in leakage beyond Technical' Specification limits. On PWRs, the use of new or existing valves which are larger than the LOCA defini-tion will require the addition of a block valve which can be closed remotely to terminate the LOCA resulting from the inadvertent opening of the vent.
PGE Response The Trojan Head Vent System is sized with a 3/8-in. dia-meter restricting orifice to limit the flow to less than 10 CFR 50, Appendix A, LOCA criteria.
Thus, no block valves are required in the Head Vent System.
The pres-surizer PORVS are equipped with motor-operated block valves which can be closed from the control room to terminate the inadvertent opening of the PORVs.
4.
An indication of valve position should be provided in the control room.
PGE Response Control room valve position indication has bean included in the Trojan Head Vent System design.
Stem-mounted limit switches are provided on the solenoid operated vent valves to indicate a position of each valve in the control room.
The manual valve located upstream of the head vent valves is kept in a locked open position and no valve position indication is required. The position of pressurizer PORVs is also indicated from stem-mounted limit switches in accordance with NUREG-0578, Section 2.1.3.a.
5.
Each vent should be remotely operated from the control rocm.
PCE Response The above Reactor Vessel Head Vent System solenoid-operated valves, the pressurizer PORVs, and the pres-surizer PORV motor-operated block valves are all operable from the control room.
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6.
Each vent should be seismically qualified.
PCE Response i
All of the piping and components for the Reactor Vessel Head Vent System and the pressurizer PORV piping will be seismically qualified.
7.
The requirements for a safety grade system is the same as the safety grade requirements on other Short-Term Lessons Learned items; that is, it should have the same qualifica-tions as were accepted for the Reactor Protection System when the plant was licensed.
The exception to this requirement is that we do not require redundant valves at each venting location.
Each vent must have its power supply from an emergency bus.
A degree of redundancy should be provided by powering different vents from different emergency buses.
PCE Response The Trojan Reactor Vessel Head Vent System has been designed to the same criteria as for the Reactor Protec-tion System when the plant was licensed. The Reactor Vessel Head Vent System and the pressurizer vent piping have redundant valves powered from separate emergency buses.
8.
For systems where a block valve is required, the block valve should have the same qualifications as the vent.
PCE Response The above Reactor Vessel Head Vent System is designed with restricting 3/8-in. orifices; thus, it does not require block valves. The pressurizer PORV block valves are safety grade and powered from Class IE emergency buses.
9.
Since the RCS Vent System will be part of the Reactor Coolant System boundary, efforts should be made to mini-mize the probability of an inadvertent actuation of the system. Removing power from the vents is one step in this direction. Other steps are also encouraged.
PGE Response Prevention of inadvertent actuation of the Head Vent System will be achieved by:
- 1) physical removal of power supply; 2) key lock hand switches for operation of the valves; and 3) appropriate operating procedures. -
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e 13.
Since the generation of large quantities of noncondensible gases could be associated with substantial core damage, venting to atmosphere is unacceptable because of the associated released radioactivity. Venting into Contain-ment is the only presently available alternative. Within Containment those areas which provide good mixing with Containment air are preferred.
In addition, areas which provide for maximum cooling of the vented gas are pre-ferred. Therefore, the selection of a location for vent-ing should take advantage of existing ventilation and d
heat removal systems.
PCE Response The Reactor Vessel Head Vent System is designed such that noncondensible gases will be discharged over the refueling cavity in order to obtain good mixing and cooling from the polar cooler fan discharge.
11.
The inadvertent opening of an RCS vent must be addressed.
For vents' smaller than the LOCA definition, leakage detec-tion must be sufficient to identify the leakage. For vents larger than the LOCA definition, an analysis is required to demonstrate compliance with 10 CFR 50.46.
PGE Response i
Each vent path has two valves in series to minimize inad-4 vertent opening. Fluid discharge from the head vent will be collected in the refueling cavity to drain into the Containment sump whose water level is continuously moni-tored for leakage. Discharge during operation would also be detected by the increased indication of Containment temperature, pressure and radiation. Pressurizer PORV leakage would be detected by increase in pressurizer relief tank water level, pressure and temperature indications.
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Section 2.2.1.b - Shift Technical Advisors The main duty and responsibility of the Shift Technical Advisor (STA) is to provide the accident assessment function as well as the operational assessment function.
In the event of an accident, the STA will be avail-able in the control room to advise the Shift Supervisor with appropriate accident response actions.
Since in a long-term modification the Techni-cal Support Center (TSC) will be equipped with data transmission system for necessary Plant parameters, the STA will not be engaged as the primary communication point between the control room and the TSC, but will be a technical advisor to the operations personnel.
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Section 2.2.2.a - Control Room Access The current Trojan plant Administrative Order A0-3-8, " Control Room Operations" specifies the limited control room access as follows:
"The Control Operator is authorized to refuse entry, or to direct person-nel to leave the Control Room if their presence interferes with plant operations or compromises plant safety, although his decision may be overruled by the Shif t Supervisor."
Based on the concerns raised by members of the NRC staff on availability of the Control Operator, PGE agrees with the NRC staff position to modify the current A0-3-8.
The modified Administrative Order will specify that the Shif t Supervisor will be responsible and authorized to refuse entry or to direct personnel to leave the control room, although, if necessary, the Shift Supevisor may direct the Control Operator to carry out this responsibility and authority for the limited access to the control room.
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