ML19260C712
| ML19260C712 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/02/1980 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578 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, TAC-30693, NUDOCS 8001080488 | |
| Download: ML19260C712 (43) | |
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January 2,1980 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Sir:
Responses to the NRC letters of September 13 and October 30, 1979, concerning implementation of NUREG-0578 short-term Lessons Learned action items, were submitted on October 17 and November 20, respec-tively. Supplementary informatica w;s submitted on December 7, December 14 and December 20 following additional reviews of Trojan design.
In the PGE responses, and in telephone conversations with members of your Staff, commitments were made to provide additional information regarding the design details of those modifications to be completed by January 1, 1980 and not requiring prior NRC approval, as well as schedule information of NUREG-0578 short-term Lessons Learned action items required to be completed by January 1,1981.
The attached report provides this information.
Sincerely, CY C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation and Maintenance CG/KM/mg/4 sala 10 Attachment h,g @
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Mr. Lynn Frank, Director g
State of Oregon 7
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Department of Energy X\\ (7
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Mr. A. Schwencer Operating Reactors Branch #1 Division of Operating Reactors U. S. Nuclear Regulatory Commission 8001080488 1703 233
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ATTACHMENT 1 TROJAN NUCLEAR PLANT Supplemental Information on NUREG-0578 Lessons '. earned Short-Term Actions 1703 234
3 Section 2.1.1 - Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs As indicated in detail in PGE Reponse to the NRC dated October 17, 1979, current Trojan design for emergency power supply satisfies the NRC requirements and no equipment changes are necessary.
Changes in the Emergency Instruction for manual connection of pressurizer heaters to the emergency buses and associated operator training have been completed and placed into effect by January 1,1980.
With the new revi-sion of the Emergency Instruction EI-4 " Plant Operation After a Loss of Normal and Preferred Power", the operator is required to manually connect at least 150 kW of heater capacity to an emergency bus if a loss of off-site power cannot be restored to the pressurizer heaters within 60 min.
The following is additional information on the Trojan design in regard to the CLARIFICATIONS provided in the NRC Letter, dated October 30, 1979.
Concerning Item 7 in the CLARIFICATION SECTION, our proposed method (October 17, 1979 PGE Reponse) for implementing this requirement does not call for automatic shedding of the pressurizer heaters upon occurrence of a safety injection (SI) actuation signal, once the heaters are connected to the 4-kV emergency diesel generator bus.
By Trojan design, the pres-surizer heaters are normally fed from a non-ESF bus but are accessible to the diesel via manual operator action. In accordance with our FSAR com-mitments (Section 8.3), the non-ESF buses are isolated from ESF buses by Class IE isolation breakers. Load shedding, by oue design, is initiated by under voltage occurring at the 4-kV bus.
The diesel generator capacity is not exceeded by the pressurizer heater load when added to the sum total of other ESF loads. The generators are each rated 4418 kW (annual con-tinuous rating); the maximum output required for each generator from FSAR Table 8.3-4 is 3364 kW, excluding the pressurizer heater load. Diesel generator starting and loading capability is not affected by the addition of 150-kW pressurizer heater loci. The pressurizer heaters are shed along with other non-ESF loads on loss if offsite power and are reconnected to the ESF bus only by emergency procedures.
With regard to the Clarificaticn Section, Item 4, instrument air is required for the PORVs and is supplied by Seismic Category I air accumu-lators. As discussed in our response dated October 17, 1979, the accumu-lator tanks are sized to support approximately 30 PORV operations and no emergency power supply is required for this operation.
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Section 2.1.2 - Performance Testing for Relief and Safety Valves A program for safety and relief valve testing has been developed by EPRI. A description of this program entitled " Program Plan for the Performance Verification of PWR Safety / Relief Valves in Systems, December 13, 1979" was submitted to the NRC on December 17, 1979 (letters from W. J. Cahill, Jr., Chairman of EPRI Safety and Analysis Task Force to H. Denton and D. Eisenhut, NRC). Portland General Electric Company (PGE) considers the program to be responsive to the requirements set forth in Item 2.1.2 of NUREG-0578. The EPRI program plan provides for completion of the essential portions of the test program by July 1981. PGE will be participating in the EPRI program to provide program review and to supply plant-specific data as required. CJP/4sa10.lA4 1703 236
Section 2.1.3.a - Direct Indication of PORV and Safety Valve Poeition Pressurizer Safety Valve Indication: In accordance with our response provided in PGE Letter dated October 17, 1979, the three safety valves have been provided with reliable flow indication by means of accoustic monitoring devices installed on each safety valve discharge line. The entire monitoring package consists ot three separate field mounted channels of flow sensing equipment (one for each valve) with individual flow indicating modules, a common power supply and a common output module mounted in an instrument bin which is located in control panel C-12 (see attached Figure 1). Each separate channel consists of an accelerometer banded to the dis-charge of the safety valves. The accelerometer is sensitive to frequen-cies that are known to be generated through the full range of valve flow. The signal generated by the accelerometer is then passed to the charge converter. The charge converter is a signal conditioning device that converts accelerometer output into a signal suitable for transmission to the control room. The equipment mentioned so far is mounted in the Containment near the top of the pressurizer enclosure. From the charge converter the signal is transmitted to the control room via channelized conduit and raceway. Each channel has an indicating module in the control room with ten indicating lights to show flow from 1 percent to 100 per-cent. At a preset level (currently 25 percent), the indicating module passes an alarm signal to the relay module which in turn triggers an annunciator on the main control board. To ensure a " safety grade" install-ation, the entire system was treated as safety related and installed as such with appropriate quality assurance procedures and documentation. Power is supplied from a battery-backed vital instrument bus and all cable and raceway is channelized. We feel that the installation, as described above, is sufficient to provide the operator with a reliable system of positive valve position based on accurate flow indication. The new system, in conjunction with previously existing indirect valve position indication described in our October 17, 1979 response, fully meets the requirements of the safety valve portion of NUREG-0578. Pressurizer Relief Valve Indication: Pressurizer PORV limit swit hes have been upgraded for post-LOCA environ-mental conditions (in accordance with the Trojan FSAR). Pursuant to the PGE response dated December 7,1979, an alarm (annunciator window) has been provided in the control room. As required, it is designed to alarm whenever either of the valves is not fully closed. )[Q} }}/ HEW /KM/sa/4gah9814.
CONTAINMENT PENETR ATION S i g l S I c> n l k SAFETY VALVES t i I i 4 r--- ACCELEROMETElk. AND I )] l BATTERY BACKED 120V 3 MOUNT I POWER SUPPLY r f l n l N l l 'I I t i INSTRUMENT BIN E 2 l )>1oe.l___ o 4-f l CHARGE CONVERTER l PSZR I n i I SAFETY RELIEF Q l g \\ [J l l VALVE LEAKAGE I I HIGH l l I L _ _ _ _ _ _ _ _ _ _ _ _ CON TR O L ROOM -4, i i PSZR ENCLOSURE + riniire i PRESSURIZER SAFETY VALVE ACOUSTIC FLOW MONITOR SYSTEM
Section 2.. 3.b - Instrumentation for Detection of Inadequate Core Cooling 1. The Westinghouses Owners' Group, of whi'ch PCE is a member, has performed analyses in response to Item 2.1.9 to study the effects of inadequate core cooling (ICC). A discussion of these analyses was submitted to the NRC Bulletins and Orders Task Force for review on October 31, 1979. As part of the Owners' Group submittal, an " Instruction to Restore Core Cooling During a Small LOCA" was included. This instruction forms the basis for procedure changes and operator training required to recognize the existence of ICC and restore core cooling based on existing instrumentation. PGE has incorporated the key considerations of this instruction into the Trojan Emergency Instruction EI-2, and has provided training to the operators in this area. Subcooling Margin Monitors Reactor coolant subcooling margin monitors (SMMs) have been installed to provide on-line indication of the saturation condition of the core and reactor coolant loops. The temperature signals are obtained from RTDs in each hot and cold leg of each coolant loop and from selected in-core thermo-couples (T/Cs). The pressure signals are obtained from two wide-range sensors located in the RHR suction leg of reactor coolant loop 4. Indication is provided at the electronics drawer. on control room panels C09A and C09B and at main control board C12. The panel C12 meters will show margin to saturation pressure in psi or temperature above saturation in *F, whichever condition exists. A switch is provided at each meter to give the operator an option to read calculated margin based on either RTD or T/C temperature sensor. At the digital meters on the electronics drawer display panels, the operator may select saturation margin, T at, P s sats AT in loops 1 and 2, at in loops 3 and 4, and each temperature and pressure input. Two levels of alarm are provided to indicate saturation conditions. The lower level, or caution, will be indicated by yellow tempera-ture sensor status lights on the display panel and actuation of the Plant annunciator. The setpoint for this alarm is 55'F below saturation. The higher level alarm will be indicated by red temperature sensor status lights on the display panel and actuation of the Plant annunciator. The setpoint for this alarm is 25*F below saturation. A built-in SMM test system is actuated manually by a selector switch. Error messages will automatically be displayed on the panel for the following conditions: all pressure inputs disabled, all RTD inputa disabled, all thermocouple inputs disabled, both RTDs and T/Cs disabled, Tref in the refertuce junction box for T/Cs out of range, and pressure above critical pressure. 1703 239.
The temperature and pressure signals for each indication channel are transmitted from their measurement loops to the margin monitors in separated raceways. The electronics drawers for each channel are separated by a r, teel barrier in panel C09. Table 1 provides detciled information of the SHMs as requested in the NRC letter dated Cetober 30, 1979. 2. The Vestinghouse Owners' Group submittal referenced above includes a discussion of the use of the core exit thermocouples for deter-mining the existence of ICC conditions. It was concluded that, in some instances, the thermocouples are superior to the loop RTDs for measuring true core conditions. Other possible means of determin-ing the approach to or existence of ICC are discussed below. Incore Detectors The use of the movable incore detector to determine the existence of ICC does not appear feasible. The detectors could be driven in to the top of the incore thimbles, which are located at the top of the core, during an accident when ICC is 'a concern. Howeve r, these detectors lack the required sensitivity at low neutron levels to detect the changes in neutron level that would occur due to core uncovery. Gamma detectors would be subject to the same sensitivity limitations, particularly since there would be a very small differ-ence in the fuel region gamma levels between the covered and uncovered conditions. It does not appear worthwhile to further pursue the use of incore detectors for detecting ICC conditions. Excore Detectors The only excore detectors suitable for detecting ICC are the source range monitors since the intermediate range and the power range monitors are not sensitive enough to detect the small neutron level changes resulting from vessel voiding. It is expected that the use of the source range monitors is limited to instances where significant voiding exists in the vessel downcomer, since water in the downcomer would ef fectively shield the source range monitors from the core region whether or not voids exist in the core region. The use of the source range monitors will be further invesrigated in connection with the continuing ICC studies described in the response to Item 2.1.9. Reactor Coolant Pump Motor Current Reactor coolant pump (RCP) motor current fluctuations could be indicative of void formation in the RCS and hence ICC conditions. These indications would, of course, not be available if the RCPs are tripped during a transient. Steam Generator Pressure Changes in steam generator secondary side pressure could be of some use in monitoring for the interruption of natural circulation h .t 1703 240
transfer from primary to secondary. However, this parameter does not directly monitor true core conditiera or indicate the approach to ICC. Reactor Vessel Level Indication A direct measurement of reactor vessel water level appears to be a possible means to provide additional capability of determining the approach to.and existence of ICC. Several systems for determining vessel water level are currently under review by the Westinghouse Owners' Group, and a final design has not yet been developed. A conceptual design for one such system is shown in Figure 2. This design is based on a differential pressure measurement between the bottom and the top of the reactor vessel. The lower sensing line of the instrument is connected to a conduit of the in-core detector system, either at the seal table or in the conduit below the vessel. The sensing line includes an isolation valve and is connected via a hydraulic coupler to a sealed reference leg. The upper sensing line is connected either to the existing vent which is to be used for the head vent system or to a spare RCCA mechanism penetration. The upper sensing line is also connected via an isolation valve and hydraulic coupler to a sealed reference leg. The differential pressure transmitter is located above the maximum Containment flooding level. The usefulness of t;iis instrument in providing unambiguous indica-tion to the operator of the ICC depends on the behavior of the instrument output generated during normal and accident conditions. These conditions, and the potential errors and accuracy of the system are being further evaluated as part of the ongoing study of ICC in connection with Item 2.1.9. 1703 241 CJP/KM/4sa10.lA15.
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w.a Reactor l l Vessel Seal l_ _ _ _ _ _ _ j Table Pmtective Incore Environment Conduit Containment Wall yR y-V l Alter: rte N oo [ Connection (Bottom) Figure 2 Reactor Vessel Nw Level Instmmentation N g 6 u -e g
TABLE 1 INFORMATION REQUIRED ON THE SUBC00 LING METER Display Information Displayed (T-TSAT, TSAT, Press, Etc) - Saturation margin, TSAT. PSAT, TCORE-THL, TCORE-TCL, individual sensor readings. Display Type ( Analog, Digital, CRT) - Analog and Digital. Continuous or on Demand - Analog is continuous, Digital is on demand. Single or Redundant Display - Redundant. Location of Display - Analog is on Panel C12 in the control room; Digital is eu new Panel C09 in the control room. Alarms (include setpoints) " Caution" at 55*F and " Alarm" at 25*F below saturation. Overall Uncertainty (*F, psi) - N/A. Range of Display - Digital (Engineering Units) 3 Digits for RTDs 4 Digits for others Analog 0-1000 psi margin to saturation 0-2000*F superheat Qualifications (seismic, environmental, IEEE-323) - Later. Calculator Type (process computer, dedicated digital or analog calculator - Dedicated Digital. If process computer is used, specify availability (percent of time) - N/A. Single or Redundant Calculators - Redundant. Selection Logic (highest temp., lowest press.) - Highest temperature *, lowest pressure. Qualifications (seismic, environmental, IEEE-323) - Later. Switch selection of highest cold / hot leg temperature or highest incore T/C. } }d}.
TABLE 1 Calculator (Continued) Calculational Technique (steam tables, functional Lit, ranges) - 4th order polynomial conversion, calculated steam table data to 0.5'F accuracy. Input Temperature (RTDs or T/Cs) - RTDs on hot / cold legs, incore T/Cs. Temperature (number of sensors and locations) - 4 each hot and cold leg RTDs, 16 incore T/Cs. Range of temperature sensors - RTDs 0-700*F, T/C, 0-2300*F. Uncertainty ** of temperature sensors (*F at 1) - N/A. Qualifications (seismic, environmental, IEEE-323) - Later. Pressure (specify instrument used) - Later. Pressure (number of sensors and locations) - 2, RHR Suction Loop 4. Range of ?ressure Sensors 3000 psig. Uncertainty ** of Pressure Sensors (psi at 1) - N/A. Qualifiestions (seismic, environmental, IEEE-323) - Later. Backup Capability Availability of Temperature and Pressure - Available by existing indications. Availability of Steam Tables, Etc. - Pressure-temperature curve in the emergency procedure. Training of Operators - Operator training relative to the revised Emergency Instructions has been completed as stated in Section 2.1.3.b. Procedures - Modifications in Emergency Instructions have been completed (see Section 2.1.3.b).
- Uncertainties must address conditions of forced flow and natural circulation.
CCT/mg/4sa10.lA18 1703 244
Section 2.1.4 - Containment Isolation The design of the Trojan Containment Isolation System and its compli-ance to the NRC requirements for NUREG-0578 Positions 1, 2 and 3 were discussed in detail in our submittal dated October 17, 1979. With regard to the NRC Position 4, necessary modifications have been completed, in accordance with the PGE letter from C. Goodwin, Jr. to A. Schwencer (NRC), dated December 20, 1979, on 23 Containment isolation valves to preclude a valve opening after a Containment Isolation Signal is reset. The following is a list of 23 valves whose control circuits have been modified: Containment sump pump line; CV-4181 and MO-4180 (FSAR Figure 6.2-10) Steam generator blowdown sample line; CV-2809, -2811, -2814 and -2880 (FSAR Figure 6.2-25) Letdown isolation line; CV-8149A, B and C (FSAR Figure 6.2-2) Accummulator tank common sample line; CV-5652 (FSAR Figure 6.2-26) Reactor Coolant System hotleg sample line; CV-5655 (FSAR Figure 6.2-36) Pressurizer liquid space sample line; CV-5657 (FSAR Figure 6.2-35) Pressurizer vapor space sample line; CV i659 (FSAR Figure 6.2-1) Reactor coolant drain tank sample line; CV-5661 (FSAR Figure 6.2-11) Reactor coolant drain tank outlet; CV-4006 (FSAR Figure 6.2-13) Reactor coolant drain tank nitrogen; CV-4000 (FSAR Figure 6.2-47) Gas collection header; CV-4301 (FSAR Figure 6.2-37) Service air to Containment; CV-4470 and -4471 (FSAR Figure 6.2-15) Containment purge supply line; CV-10001 (FSAR Figure 6.2-46) Containment purge exhaust line; CV-10004 (FSAR Figure 6.2-48) Chilled water return line; CV-10014 (FSAR Figure 6.2-41) Chilled water supply line; CV-10015 (FSAR Figure 6.2-40) The 23 Containment isolation solenoid valve circuits have been modified in one of two ways. In the first case, the existing control switch con-tact configuration was modified and used with a new auxiliary relay. The auxiliary relay is energized through the control switch and the normally i703 245
closed Containment isolation signal contact to pick up the solenoid valve coil causing the valve to open. A contact off of the auxiliary relay seals it in the circuit. On the other hand, the auxiliary relay is deenergized, closing the valve by either a control switch operation or a Containment isolation signal. Deenergizing the auxiliary relay breaks the seal-in circuit. In the second case, the existing control switch contact configuration was modified and a spare limit switch contact utilized to seal-in the solenoid valve coil when the valve is fully open. Upon receipt of a Containment isolation signal or actuation of the control switch, the solenoid valve coil is deenergized causing the valve to close. The limit switch contact opens to break the seal-in when the valve begics to close. In both these modifications, manual operator action of the control switch is required to reopen the solenoid valves following a Containment isolation signal reset, since the seal-in contact has opened to prevent the solenoid valve coil from reenergizing. KM/ RAY /4 sala 2 1703 246
Section 2.1.5.a - Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems As described in the PGE response dated October 17, 1979, the Trojan plant design currently satisfies the h1C requirements and thus no action is required on this item. KM/4sa10A30 1703 247
Section 2.1.5.c - Hydrogen Recombiners As described in the PGE response dated October 17, 1979, the Trojan plant design currently satisfies the h1C requirements and thus no change has been made on this item. 1703 248 KM/4sa10B1.
2.1.6.a - Integrity of Systems Containing Radioactive Fluids The following leakage rates have been determined for systems containing liquids: Safety Injection System Nondetectable CVCS 200 cc/ day Residual Heat Removal 50 drops / day Containment Spray Nondetectable These leak rates were estimated from a review of existing Trojan Nuclear Plant Maintenance Requests which were generated as a result of the existing preventative maintenance program for keeping leakage from radioactive systems to as low as practicable levels. The total leak rate from the radioactive gaseous waste svstem has been estimated to be 20 cc/ min. This leak rate was determined from the calcu-lation of maximum leakage that could be present while not being detected by Auxiliary Building airborne radiation sampling. Specific procedures to perform periodic leak tests of these systems have been prepared. Tests will be completed prior to Cycle 3 startup in accordance with our commitment in our letter dated October 17, 1979. In the course of our review requested by NRC IE Circular 79-21 and NRC letter dated October 17, 1979, we have identified a potential path for uncontrolled release of radioactive gas and/or liquid: the refueling water storage tank (RWST) overflow line ccnnects to a common drain header going to the clean waste receiver tanks. The drain header shown on FSAR Figure 11.2-9 collects liquid waste from radioactive systems in the Auxiliary and Fuel Buildings. The line is also vented to the atmosphere at the RWST. Since most of the other drains in the header contain radioactive liquid, this path could permit an unplanned release. A design change has been written to prevent uncontrolled leakage from this system. Similarly, the floor drains in the Auxiliary Building connect into a common drain header prior to going to the dirty radwaste system. Also connected to this system are five open floor drains in the uncontrolled Turbine Building facade area. Cross connections of these drains could cause liquid or gaseous releases to the environment. The drainc are temporarily capped and a design change has been written to reroute the drain. Since each drainage system, including the RWST overflow line, must be analyzed and resolved separately, engineering on these modifications will extend over the first quarter of 1980. Procurement entails approximately 6 months and installation should begin by the fourth quarter of 1980. We are continuing our review of potential paths for uncontrolled release of radioactivity. Potential paths identified in this continuing review will be reported to the NRC. 1703 249 GRC/mg/4 sala 9 Section 2.1.6.b - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May be Used in Post-Accident Operations A design review of radiation shielding at Trojan has been completed for vital areas and safety equipment locations to ensure that personnel occupancy will not be unduly limited by post-accident radiation fields. The methodology and results of the design review are described below. METHODOLOGY The review was performed in the following manner: a. Po s t-accide r". operating conditions were determined to identify piping which could potentially contain highly radioactive liquids and gases, b. Source terms were determined using NRC position clarifi-cation 2.1.6.b in Enclosure 2 of Mr. H. Denton's letter dated October 30, 1979. c. Locations within the Plant where operators were required to perform specific functions were identified. In this review, the assumption was made that equipment operated as designed. d. Dose rates were determined at locations identified in [c] based on the operating conditions and source terms developed in [a] and [b]. POST-ACCIDENT OPERATING CONDITIONS Accident conditions considered both the post-accident conditions (eg, main steamline break, LOCA, feedwater line break, etc) described in Trojan FSAR Section 15.4 and a nonmechanistic accident resulting in high levels of radioactivity in systems used to mitigate accidente. The above considerations were reduced to two limiting conditions:
- 1) the design basis LOCA with complete severence of reactor coolant pipe (Post-LOCA Fode); or 2) a secondary or primary side initiated event with the Reactor Coolant System (RCS) remaining intact (Intact RCS Mode).
In both cases, extensive core damage with coincident release of fission products from the fuel was assumed. Depending on the accident, this radioactivity was released to either the contained Reactor Coolant System or to the Containment sump. The Intact RCS Mode assumed nonmechanistic hydrogen production requiring degassification. A. Post-LOCA Mode Description - The Post-LOCA Mode assumed a large break Loss-of-Coolant Accident. All ECCS pumps operated and within 30 minutes were circulating Containment recirculation sump water through the Auxiliary Building pip-ing. Systems assumed to operate during this period included RHR in the recirulation mode, safety injection drawing from 1703 250.
RER, Containment spray in the recirculation mode, component cooling watcr, service water, hydrogen monitoring and Reactor Coolant System sampling. Recirculation from the Containment sump was assumed to continue indefinitely. B. Intact RCS Mode Description - The Intact RCS Mode was developed to bracket all other major accidents. Addition-ally, a nonmechanistic fuel failure with hydrogen production was assumed which required RCS degassing within 30 minutes to evaluate dose rates from the Waste Gas System. This mode assumed all ECCS pumps operating; however, only non-radioactive water was assumed to be initially injected into the core. The ECCS repressurized the RCS and subsequent core cooling was accomplished by natural circulation. Depressurization would have resulted in the Post-LOCA Mode described above. Systems assumed to operate during this mode include: a. Chemical Volume and Control System (CVCS) letdown to the volume control tank, and CVCS charging back to the RCS from the volume control tank for RCS degassing. b. Reactor coolant pump seal injection was assumed isolated immediately upon initiation of the accident. During periods of Reactor coolart pump operation, seal injection was provided from a low-activity source such as the boric acid storage ta nks. Letdown, during this condition, was valved to the CVCS holdup tanks. Degassing and reactor coolant pump operation were not assumed to operate simultaneously. c. The RsS remained on natural circulation to allow decay of short-lived isotopes before switching on to RRR. At least 7 days' decay prior to going to RHR was assumed. d. Degassing of the RCS was accomplished by letdown to the volume control tank which is supplied with continuous nitrogen purge vented to the waste gas compressors. The compressors either discharge to the waste gas decay tanks or discharge directly to the Containment. SOURCE TERMS All source terms were based on NRC position clarification 2.1.6.b. Source terms for the Post-LOCA and the Intact RCS Modes are presented below. A. Post-LOCA Mode Source Term - The liquid source term con-sisted of Containment sump water containing 50 percent halogens, 1 percent other nuclides and a negligible amount 1703 251
of noble gases, diluted with accumulator and refueling water storage tank inventories. Noble gases present initially in the water are rapidly released to the Contain-ment atmosphere. Thirty minute decay was assumed based on the approximate time required to switch over from injection from the RWST to recirculation from the Containment sump. The hydrogen monitoring system source assumed 100 percent noble gas and 25 percent halogens diluted into the net free volume of the Trojan Containment. B. Intact RCS Mode Source Term - The liquid source term con-sisted of 100 percent noble gas, 50 percent halogen, and 1 percent other nuclides, undiluted and decayed 30 minutes to accommodate expected times before operator presence was required. The noble gas source term in gas-containing lines between the volume control tank and the waste gas compressors was assumed to be seven times greater than the RCS concentration. This is based on the calculated maximum concentration in the VCT vapor space assuming 30 standard cu ft per minute (scfm) N2 Purge rate. Four and one-half hour decay was assumed based on calculated time of maximum noble gas concentration in the VCT given a nitrogen purge rate of 30 scfm. The gaseous source term downstream of the waste gas compressors to the vaste gas decay tanks considered compression of the gases which would occur in the waste gas compressors. The source in the waste gas decay tanks consisted of approximately 50 percent of the RCS liquid noble gas inventory with 2 days' decay; ie, all RCS noble gas was assumed to be contained within two waste gas decay tanks. Two days' decay was based on the theoretical time required to degas the RCS. Hydrogen monitoring source term was assumed to be identical to the Post-LOCA Mode described above. PERSONNEL OCCUPANCY REQUIREMENTS Personnel occupancy requirements were developed based on post-accident operating modes discussed above. A. Post-LOCA Mode Occupancy Requirements - Since all ECCS equipment is redundant and remotely operated, operator access to pumps / valves etc are generally not required in that mode. Access was assumed to be required to the components listed below: a. Component cooling water system valve lineup; b. RCS sampling; 1703 252.
c. Radwaste control panel for monitoring conditions in the Auxiliary Building; d. Hydrogen monitoring cabinets; and e. Containment and Auxiliary Building Ef fluent Radia-tion Monitors. Access to the hydrogen monitoring cabinets was assumed to be required within 30 min of the start of the accident. B. Intact RCS Mode Occupancy Requirements - Occupancy require-ments for the Intact RCS Mode included the occupancy requirenent described in the Post-LOCA Mode described above plus: a. Access to the waste gas compressors prior to degassing; ,b. Access to the waste gas storage tank drain valves during degassing; c. Ability to line up the waste gas compressors to the Containment prior to degassing; d. Manual operation of the hydrogen and nitrogen purge valves to the volume control tank during degassing; and DOSE RATE ACCESS CRITERIA The criteria used assumed that the dose rate should be less than 5 rem /hr for areas ree tiring access less than 10 min, less than 1 rem /hr for areas requiring adcess of 10 min several times per shif t, and less than 15 mrem /hr for areas requiring continuous access. The access times were determined based on review of operating procedures and discussion with operating personnel. DOSE RATE DETERMINATION Graphs of dose rate versus distance from various diameter pipes, attenu-ation factors, and decay factors for each source term were calculated using the ISOSHLD computer code. ISOSHLD is point kernel shielding code utilizing the linked fission product inventory code, RIBD. PGE's version of ISOSHLD was developed from the Battelle Northwest shielding code, BNWL-236. All piping and locations requiring occupancy identified in the above operating conditions were marked onto piping layout drawings. The pipes and equipment in the vicinity of each location requiring occupancy were identified, the distance to the dose point, diameter, source term and thickness of concrete shielding were tabulated, and the appropriate dose contribution from each source point was evaluated. All significant 1703 253.
sources were added to give the total dose contribution at the dose point. Total dose rate was then compared with radiation zone criteria. Access pathways were also evaluated to assure accessibility to the required occupancy locations during post-accident conditions. RESULTS Potential modification to shielding, equipment design or operating pro-cedures was identified in cases where the evaluated dose rate exceeded the radiation zone levels. A detailed evaluation of each potential modification will be conducted and the required modifications and procedure changes will be completed prior to January 1, 1981. At this point, modifications that may involve additional shielding include: a. The radiation sample room which is addressed in Response to Item 2.1.8.a); b. The hydrogen monitoring cabinet; c. Pipe tunnel over the corridor at Elevation 77 ft near the radwaste control panel; and d. PRMS-1 and 2 at Elevation 93 ft. Components which may require reach rods or remote operators include: a. RHR low pressure letdown path manual valve 8734A and B. b. Capability to isolate seal injection flow to one reactor coolant pump remotely. c. Isolation / bypass of the reactor coolant filter. Air-operated valves which may require accumulators or remote air hookups include: a. Letdown orifice isolation valves (CV8149A-C). b. Flow control valve F110 and Flll boric acid blender. Components which may require modification include: The seal injection filter cartridge to allow remote handling by the existing filter handling machine. Procedures which may require modification or development include: a. Capability to provide reactor coolant pump seal injection f rom the boric acid storage tank during periods of high activity reactor coolant. 1703 254.
b. A procedure which would allow delay of RHR initiation for periods when high activity reactor coolaat is present and no Loss-of-Coolant Accident has occurred. c. Capability to degas the reactor coolant system without cycling the volume control tank. The current procedure, if not chang-d, may require modifi-cation of the waste gas surge tank shielding, addition of remote operators on the WGST drain valve, addition of accumulators on the volume control tank gas collection header isolation valve CV-8102, and modification of N2 inlet and gas collection header vent connections to the volume control tank. Alternate degassing procedures :rhich may preclude the need for the above potential modifications are being investigated.
SUMMARY
In summary, the design review of radiation shielding at Trojan has con-firmed that the control room and proposed technical support center are not af fected by increased source terms. Several areas have been identi~ fied which require further evaluation to determine whether additional shielding, remote operators, air-operated accumulators or procedure changes are required to ensure that personnel occupancy will not be unduly limited by post-accident radiation fields. Modifications to shielding, equipment or procedures will be completed by January 1, 1981. 1703 255 JLT/4mg10A26.
t Section 2.1.7.a - Automatic Initiation of Auxiliary Feedwater System The design description of the Auxiliary Feedwater System automatic initi-ation was originally provided to the NRC in PGE response on October 17, 1979. Subsequent to this submittal, supplemental information was pro-vided on November 26 and December 14, 1979 in response to the NRC letters Request for Information dated October 3 and December 10, 1979, respectively. It is our belief that the above submittals provide suf-ficiently descriptive information to the NRC Staff members for an inde-pendent review of this item. As described in the above responses, the automatic initiation signals and circuits of the Trojan Plant Auxiliary Feedwater System presently meet all NRC requirements, thus no change has been made. \\ 7 03 2.Su6 KM/4sa10B2 _22_
Section 2.1.7.b - Auxiliary Feedwater Flow Indication to Steam Generators The design description of the Auxiliary Feedwater System automatic ini-tiation was originally provided to the NRC in a PGE response on October 17, 1979. Subsequent to this submittal, supplemental information was pro-vided on November 26 and December 14, 1979 in response to the NRC letters Request for Info' mation dated October 3 and December 10, 1979, respec-r tively. It is our belief that the above submittals provide sufficiently descriptive information to the NAC Staff members for an independent review of this item. In accordance with the PGE commitment !n the supplemental submittal dated December 14, 1979, the design change ti a been made and the auxiliary feedwater flow instrument channels are presently powered f rom the vital instrument buses. The design of the existing cables for the AFW flow indicators satisfies the same requirements as for the other iadication devices in accordance with the licensing commitments in the Trojan FSAR. Therefore, by modifying the power supply to the AFW flow instrument channels, it is our belief that the Trojan design satisfies the NRC requirements in NUT.vs+0378 as well as the NRC letter of clarifications on October 30, 19??. i703 257 KM/4sa10B3.
Section 2.1.8.a - Improved Post-Accident Sampling Capability 1. A review of the current sampling and analysis procedures at the Trojan Nuclear Plant has resulted in the following procedural modifications to improve the ability to obtain and analyze post-accident samples of reactor coolant: a. Sample tongs and portable shielded containers will be used if necessary to reduce personnel exposures when obtaining reactor coolant samples. b. The in-Plant Ge-Li detector will be modified by using collimators and/or increased distance between detector and sample to allow counting of higher-activity samples. c. Chemical analysis capability will be improved by using extended electrode leads and an extended delivery tube for NaOH titrate for chloride and boron analyses, as well as portable lead shielding in the sample sink of the chemistry laboratory. These procedural modifications have been completed and the revised procedures are in place at the Trojan plant. 2. A shielded sampling f acility to obtain a pressurized sample of reactor coolant liquid within 1 hr following an accident will be provided. This sampling facility will be located in or adjacent to the existing sampling room in the Auxiliary Building, Eleva-tion 45 ft, and will limit personnel exposures in accordance with NUREG-0578 requirements. Sample lines will be routed to this station from taps on the existing RCS sample lines downstream of the sample cooler. Appropriate system isolation valves will be provided. The capability to obtain a representative sample of approximately 5 milliliter volume will be provided by a return recirculation flow through the sample container back to the exist-ing recirculation flow line to the volume control tank. The sample container will be removable and sized for transport in a shielded container for analysis. Equipment for handling and transport of the shielded container will be provided. Capability to perform chemical analysis of the sample for boron and hydrogen will be developed with proper operational procedures to maintain personnel exposures as low as reasonably achievable and to meet the 3 and 18-3/4 rem dose limits to the whole body and extremities, respectively, per NUREG-0578. We are also evaluating the availability and capability of in-line instrumentation for the hydrogen and boron analyses. A radionuclide analysis will be performed on a grab sample by transporting it to a remote mobile low-background laboratory which will be equipped with a Ge-Li detector if the in-Plant counting facility cannot be used. Sample disposal following analysis may be accomplished via connections to existing Plant radwaste systems or a designated sample container with sufficient shielding. We are continuing evaluation of the 1703 258.
necessity for chloride analysis on the RCS liquid sample. Westing-house is also evaluating this issue for the Owners' Group and the final report is due in February 1980; PGE will evaluate the final report before any decision is made. 3. Samples of Containment atmosphere will be obtained by providing a tap on the existing post-accident hydrogen analysis sampling line for the purpose of routing a Containment atmosphere sample to a new sample facility to collect grab samples under positive and negative Containment pressure. A representative sample will be obtained by use of recirculation flow back to the Containment. Sample recirca-lation under both positive and negative Containment pressure can be obtained by use of a recirculation pump (this pump may be one of the existing hydrogen analysis system pumps). The new grab sample facility will conform to the seismic design and quality group classi-fication of the system to which each sampling line is connected. The grab sampling line and piping internal to the grab sampling panel, its in-line equipment (pumping and valves) and the grab sample vessel will be shielded and designed such that an individual may promptly and safely obtain a sample and not receive a radia-tion exposure in excess of " rem to the whole body or 18-3/4 rem to any extremity. A grab ssmple vessel of approximately 5 milli-liter sample volume will be provided with a shielded container so that it may be removed and transported for remote analysis in the radiation sample room or elsewhere. Adequate radiation protection will be ensured by requiring monitor-ing of radiation level from the (shielded) sample piping before and during sampling. 4. PGE will provide an additional description of design modifications currently under internal review by April 30, 1980 and intends to complete implementation by January 1,1981. Should there be any problem in material availability which may prevent meeting this date, PGE will promptly notify the NRC. 1703 259 SGG/ GEM /KM/4sa10.lA13 -2 5 -
Section 2.1.8.b - Increased Range of Radiation Monitors NRC Position 1 1. Noble Cases PGE will install high range noble gas radiation monitors on the Containment Radiation Monitoring System and the Main Condenser Air Discharge Radiation Monitoring System prior to January 1,1981. Upon installation of the high range monitors, the radiation monitoring systems for both the Containment and Main Condenser Air Discharge System will include three detectors with overlapping ranges and will provide an overall detection range of 3 x 10-7 uC1/cc (Xe-133) to 1 x 10 +5 uCi/cc (Xe-133). As agreed with NRC on December 4, 1979, PGE will modify the Auxiliary Building Radiation Monitoring System prior to startup of Cycle 3 in 1980 to increase the noble gas concentration detection range to 3 x 10+2 uC1/cc (Xe-133). This modification essentially duplicates the current Containment noble gas radiation monitoring system design for the Auxiliary Building. The modified Auxiliary Building noble gas monitoring Lystem will include two detectors with overlapping ranges and will provide an overall detection range of 3 x 10-7 uCi/cc (Xe-133) to 3 x 10+2 uC1/cc (Xe-133). These additional noble gas radiation monitors will be instrumented, recorded and alarmed in the control room and will be designed and fabricated in accordance with Section 11.4 of the Trojan FSAR. Calibration and testing of the noble gas monitors will be in accor-dance with the Technical Specifications of the Trojan Operating License. 2. Iodines PCE Response 2.1.8.b, NRC Position 2 (Page 27), addresses the procedure implemented for determining concentrations of radioiodine in effluents. In addition, existing Radiological Emergency Response Plan procedures require field monitoring for radioiodines in the event of an accident. Thus, no additional action is considered to be necessary. 3. High Range Containment Area Radiation Monitors As detailed in the October 17, 1979 letter to NRC, PCE will install high range Containment area radiation monitors in accordance with NUREG-0578 prior to January 1,1981. Should there be any problem in material availability which may prevent meeting this date, PGE will promptly notify the NRC. 1703 260 MQH/4mg10B14.
2.1.8.b Increased Range of Radiation Monitors NRC Position 2 Interim Procedures for Quantifying High Level Accidental Radioactivity Releases Interim procedures have been developed for estimating noble gas and radioiodine release rates from the Auxiliary and Fuel Buildings, Contain-ment, condenser air ejector and the steam safety and atmospheric dump valves. The interim procedures are in place at Trojan Plant and will be implemented if existing instrumentation goes off-scale. The following description of the analysis method follows the outline specified in the NRC clarification letter dated October 30, 1979. A. Noble Gases 1. System / Method
Description:
a. The in-Plant Ge-Li detector with an ND 44-20 analyzer will be used whenever possible to analyze noble gas samples. Daily calibration is done using a mixed NBS gamma standard. In the event that the in-Plant Ce-Li detector cannot be used because of high background, high sample activity or loss of power to the Ge-Li detector, the sample radioactivity will be determined using a portable survey meter. This method provides the following capabilities: 1) 1.3 uCi/cc to 105 uCi/cc Xe-133 in an air sample using an open-window Eberline E-530 survey meter with HP-270 probe. 2) 4.4 x 10-2 uCi/cc to 105 uCi/cc Xe-133 in a steam sample at normal system temperatures and pressures using the Eberline E-530 or the Eberline R0-2A survey meters. These survey meters will adequately measure gamma ray energies of 80 key and will be calibrated using a Cs-137 source in accordance with existing Plant procedures. b. The noble gas grab samples for the Auxiliary and Fuel Buildings and the Containment will be taken at the hydrogen sampling panel C-285-B at Elevation 93 ft of the Auxiliary Building. To ensure repre-sentative measurements, the gas will be recirculated in the hydrogen line. The sample will be taken to an area of low background for analysis. 17H03 261.
To quantify release rates from the steam safety valves and atmospheric steam dump valves, an undegassed liquid grab sample will be taken from the main steam sample line in the secondary chem-istry laboratory. Samples for the condensor air ejector will be taken at PRM-6 in the Turbine Building. c. Analysis will be completed by the Chemistry Department and the results transmitted to the control room using the in-Plant executone system. d. A walkthrough of the sampling procedures shows that a noble gas and iodine sample can be taken from the hydrogen sampling panel location within 6 min. This time includes time for entering and leaving the area. Analysis of the noble gas sample using the Ge-Li detector can be done in 5 min. More rapid analysis can be done using the portable survey meter. Capability, there-fore, exists to obtain radiation readings at least every 15 min during an accident. e. A-C power is used for operating the Ge-Li detec-tor. In the event that power is lost, sample analysis will be done using the portable survey meter. 2. Procedures for Conducting Measurement / Analysis: a. Occupational exposures will be minimized 'v reduc-ing time spent for obtaining the sample. . raining will be done to familiarize individuals with the sampling method in order to reduce sampling time. High activity samples will be handled remotely. The sampling location at the hydrogen sampling panel is included in the shielding analysis performed in response to Item 2.1.6.b. b. Instrument readings will be converted to effluent concentrations using equivalent Xe-133 calibration factors. c. Information will be disseminated as described in A.l.c. d. Instruments will be calibrated using procedures described in A.l.a. 1703 262.
B. Radioiodine and Particulate Effluents 1. System / Method
Description:
a. The in plant Ge-Li detector will be used when-ever possible to analyze iodine samples. Backup instrumentation including the Eberline SAM II or the VIC counting equipment descrit.ed in Response 2.1.8. is also available. In the event these instruments cannot be used because of high activity, the sample radioac-tivity will be determined using a portable survey meter. This method permits measurement of iodine concentrations up to approximately 770 uCi/cc I-131 which corresponds to the high activity source term prescribed for the shield-ing analysis in Item 2.1.6.b. b. An iodine grab sample for the Auxiliary and Fuel Buildings and the Containment will be taken at the hydrogen sampling parel C-285-B on Elevation 93 ft. The condenser air ejector sample will be taken at PRM-6 in the Turbine Building. c. Occupational exposures will be minimized by redoc-ing time spent for obtaining the grab sample. High activity samples will be handled remotely. d. If interference from noble gases is suspected, silver zeolite cartridges will be used. e. See A.l.e. Also, the Eberline SAM II is provided with batteries. 2. Procedures for Conducting Measurement / Analysis: a. See A.2.a. b. Instrument readings will be converted to effluent concentrations using equivalent I-131 calibration factors. c. Information will be disseminated as described in A.l.c. d. Instruments will be calibrated using procedures described in A.l.a. }[Q} 2h) GRC/sh/4salAl2.
2.1.8.c - Improved Iodine Instrumentation Procedures have been revised and implemented for the use of an Eberline SAM-II portable 2-channel analyzer for in plant post-accident iodine air sampling. Silver zeolite sampler cartridges will be used for collection of iodine samples if an interference from noble gases is suspected. The requirements for calibration and maintenance of the SAM-II is also addressed in the revised procedures. Procedures ddressing the use of counting equipment stored at the Visitors Information Center (VIC) as a backup for the counting of in plant air sampler cartridges have been implemented at the Trojan plant. The VIC counting facility will utilize either a 256 multi-channel analyzer with a 2 x 2 NaI detector, or a scanning single-channel analyzer with a 2 x 2 NaI detector. Procedures have been implemented for the calibration and use of both instruments; one of the detectors will be available at the VIC at all times. SCG /sa/4mglA16 1703 264
Section 2.1.9 - Analysis of Design and Off-Normal Transients and Accidents NRC Position 1 Analyses of small-break LOCA's, symptoms of inadequate core cooling, required actions to restore core cooling, and analyses of transient and accident scenarios including operator actions are being performed on a generic basis by the Westinghouse Owners' Group, of which PGE is a member. The small-break analyses have been completed and were reported in WCAP-9600, which was submitted to the NRC on June 29, 1979. Incorporated in the report were operator guidelines that were developed as a result of the small-break analyses. These guidelines have been reviewed and approved by the NRC Bulletins and Orders Task Force, and were presented to Owners' Group utility representatives at a seminar on October 16-19, 1979. Following thl= seminar, plant-specific procedures were developed by PGE for Trojan and operating personnel were trained on the new procedures. These revised procedures are Emergency Instruction, EI-0 " Safety Injec-tion and Diagnosis", EI-1 " Loss of Reactor Coolant", EI-2 " Loss of Secon-dary Coolant", and EI-3 " Steam Generator Tube Failure". The revised pro-cedures are in place and training has been completed at Trojan in accor-dance with the requirement in Enclosure 6 of Mr. Eisenhut's letter of September 13, 1979 and Enclosure 2 to Mr. Denton's letter of October 30, 1979. The work to address inadequate core cooling (ICC) is being performed in accordance with schedules and requirements established by the Bulletins and Orders Task Force. Analyses related to the definition of ICC and guidelines for recognizing the symptoms of ICC based on existing Plant instrumentation and for restoring core cooling following a smal:-break LOCA were submitted to the NRC Bulletins and Orders Task Force by the Westinghouse Owners' Group on October 31, 1979. These analysea will be followed by more extensive and detailed analyses during the first quarter of 1980. The guidelines and training based on the enalyses submitted on October 31, 1979 are in place at Trojan as required by the Bulletins and Orders Task Force. With respect to the other transients and accidents described in Chapter 15 of the Trojan FSAR, the Westinghouse Owners' Group is performing an evaluation of the actions which occur by constructing sequence-of-event trees for each of the LOCA and non-LOCA transients. From these event trees a list of decision points for operator action will be prepared, along with a list of information available to the operator at each decision point Based on the results of these efforts, criteria will be established for credible misoperation, and the time available for operator decisions will be qualitatively assessed. The information developed will then be used to test generic abnormal and emergency operating procedures against the event sequences and determine if inadequacies exist in the procedures. Following review of this work by the Owners' Group, a submittal will be made to the NRC by the end of the first quarter in 1980. This schedule is er tsistent with the requirements in Enclosure 2 of Mr. Denton's letter of October 30, 1979. 1703 265
Test prediction calculations of the LOFT L3-1 nuclear small-break experi-ment have also been completed for the Owners' Group. These calculations were submitted to the NRC on December 15, 1979 in accordance with the schedule established with the Bulletins and Orders Task Force. CJP/KM/sh/4 sala 4 ~32-j }
Section 2.1.9 - Analysis of Design and Off-Normal Transients and Accidents NRC Position 2 A design description of Containment instrumentation (Containment pres-sure, hydrogen and water level) was initially provided to the NRC in a response dated October 17, 1979 and supplemented in PGE's submittal dated November 20, 1979. As described in the above response, the current Trojan plant design complies with the NRC position ou the Containment hydrogen measurement and no modification is necessary. The existing post-accident Hydrogen Analysis System satisfies the requirements for redundancy and seismic qualification as discussed in FSAR Section 6.2.5 and illustrated in FSAR Figure 6.2-48. The hydrogen analyzer has a range of 0-10 percent hydrogen in air and is capable of running continuously through the Containment pressure range of 0-60 psig. Two new Barton dP transmitters will be installed in parallel on two f the existing four safety grade pressure transmitters which are calibrated for a range of 0-75 psig. The new transmitters, measuring Containment pressure from -10 to +190 psig, will be physically separated and will be provided with power from safety grade emergency / vital instrument buses in accordance with the Trojan FSAR. The cable runs between each new transmitter and the new control room indicator will also be safety grade. Further investigation of the means to effectively determine water level in the Containment has led to a design of one system covering both narrow-and wide-range water levels. The measurement range of the new system is from the bottom of the Containment sump (approximately Eleva-tion 41 ft) to Elevation 53 ft (equivalent to 500,000 gal of water). This measurement range is adequate to accommodate the maximum Contain-ment water volume (approximately 472,000 gal as described in Trojan FSAR Section 15.4. l.6). The bubble-tube type measuring method is under con-sideration for this application. The two new transmitters (one for each Containment sump) will be powered from vital instrument buses, and each transmitter output will be routed to a Seismic Category I indicator in the control room. Instrument loops from the transmitters to the indi-cators will be run in separate channels. There are no active instrumen-tation components for this system inside the Containment, and all active components outside of the Containment will be accessible during post-LOCA conditions. PCE will make every ef fort to complete installation of the Containment pressure and water level instruments by January 1, 1981. Should any problems arise in material availability which would prevent meeting this date, PCE will promptly notify the NRC. KM/4 sala 6 1703 267
Section 2.1.9 - Analysis of Design and Off-Normal Transients and Accidents NRC Position 3 The reactor vessel head vent system which is under consideration by PGE for the Trojan Nuclear Plant is shown in Figure 3. The system will con-sist of a 1-in. diameter vent pipe with four Safety Class 2 fail-closed isolation valves. The system will be designed for " single active failure" venting and isolation. The vent piping will be provided with a restrict-ing orifice to prevent coolant loss greater than the makeup capacity of one centrifugal charging pump. All piping and components between the orifice and the point of discharge will be designed to Seismic Category I, Safety Class 2 criteria. The solenoid isolation valves will be powered by redundant safety grade electrical supplies. The venting piping will discharge to an optimized location in Containment to provide dilution and recombination of potentially combustible gases. Safety grade components in the Containment will be protected from any discharges from the head vent system. The head vent system will be used in conjunction with a reactor vessel water level indicating system. A narrow-range vessel water level indi-cating system is being reviewed by PCE specifically for this use. This narrow-range system consists of one tap off of two hot legs, a common head vent tap and two dP cells. This vessel level system design will be reviewed in conjunction with the requirement in Section 2.1.3.d for detection of inadequate core cooling. The pressurizer power-operated relief valves (PORVs), as presently designed, provide a safety-grade capability for venting noncondensible gases from the pressurizer steam space. Power to operate the PORVs and their associated block valves is provided from safety grade circuits, as described in our October 17, 1979 response to Item 2.1.1. This design is such that venting of noncondensibles from the pressurizer would be possible even assuming a single failure of one of the two PORVs or block valves or one emergency power train. It is currently expected that the vessel head vent and associated vessel water level indicating systems will be installed and operating procedures will be developed and implemented by January 1, 1981. 1703 268 KM/sh/4 sala 7.
TO CONTAINMENT h h THic ]Hic 1 SEE NOTE 1 ORIFICES REACTOR VESSEL L, HEAD NOTES:
- 1. EXISTING VENT LINE Figure 3
~ o 1703 269. 6
Section 2.2.1.a - Shif t Supervisor's Responsibilities 1. A letter of Management Directives by the Assistant Vice President for Thermal Plant Operations and Maintenance has been issued to the Trojan plant Management and Shif t Supervisors, which emphasizes the Shif t Supervisor's management responsibilities for safe operation of the Plant and command duties which are now defined in the Trojan plant Administrative Order, A0-1-4, " Operations Responsibilities". The Management Directive will be issued annually hereafter and a copy will be kept in the Trojan plant file for review upon request. 2. A0-1-4 has been revised to clearly specify duties, responsibilities and authority of the Shif t Supervisors and control room operators for the line of command and the command-decision authority. A par-ticular emphasis is placed on the responsibility of the Shift Super-visor to maintain a broad perspective of operational conditions affecting the safety of the Plant as a matter of highest priority at all times. Based on the revised A0-3-8, " Control Room Operations", the Shift Supervisor will remain in the control room at all times during accident conditions until he is properly relieved or until the reactor coolant system is stabilized. Persons authorized to relieve the Shif t Supervisor are those holding a current senior reactor operator's license (Manager of Operations and Maintenance, and Operations Supervisor). The revised A0 also specifies duties, responsibilities and authority of the control room operator who will be designated if the Shif t Supervisor is absent from the control room during routine operations. 3. Indoctrination of the Shif t Supervisors on the revised A0s to emphasize their responsibilities and their management functions for safe operation has been held. 4. When the Management Directive was issued, a review of the admini-strative duties of the Shift Supervisor was also made to ensure the administrative functions that detract from or are subordinate to the management responsibilities for assuring safe operation of the Plant will be delegated to other personnel not on duty in the control room. KM/sh/4salAll 1703 270
Section 2.2.1.b - Shift Technical Advisor A description of the implementation program of the Shift Technical Advisor (STA) was provided to the NRC in our response dated October 17, 1979. On November 20, 1979, additional detailed information specifying two phases of the STA pregram was also submitted for your review. It is our belief that the above responses provide sufficient info rma tion on this item. As indicated in our November 20, 1979 submittal, Phase I of the STA program was implemented on January 1,1980. This Phase I program util-izes Plant Engineering personnel in the newly established pool of Plant Engineering staff and other Engineering member as the STA. The STA will be assigned onsite during Operational Modes 1, 2, 3 and 4. Opera-tional Modes 5 and 6 do not warrant the necessity of the STA since either the Safety Analyses do not consider reactor accidents during Mode 5 to be significant or significant accidents in Mode 6 are limited to fuel handling accidents during which a dedicated senior reactor operator is assigned for supervision. The shift of the STA is a 24-hr assignment and the engineer assigned will be onsite for this period. All of the engineers in the pool will be rotated on a day-to-day basis through the assignment and will be able to respond to the control room within 10 min upon notification from the operating crew through a portable pager or " beeper". A detailed description of duties and responsibilities of the STA and Phase II training program was provided in our response dated November 20, 1979. The Phase II program and training will be implemented by January 1, 1981 in accordance with the NRC requirements. KM/sh/4 sala 17 i703 271 _37_
Section 2.2.1.c - Shif t and Relief Turnover Procedures 1. The Trojan plant Administrative Order, A0-3-6 " Conduct of Opera-tions and Shift Records" has been modified to specify how the critical plant parameters are verified to be within limits; how assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transi-ents and accidents is obtained by a check of the control console; and how systems and components that are in a degraded mode of operation permitted by the Technical Specifications are tracked. For such systems and components, the length of time in the degraded mode will be compared with t'te Technical Specification action statement. Check lists were provided in the procedure for imple-mentation of these actions. The revised A0 and the check lists are in place at the Trojan plant in accordance with PGE responses to the NRC dated October 17 and November 20, 1979. 2. In order to ensure proper turnover of maintenance and testing in progress at shift change, check lists of critical equipment and components were provided for Auxiliary Building operators, Turbine Building operators and water plant operators providing shift cover-age. 3. The Plant QA audit program has been revised to require a periodic audit which evaluates the effectiveness of the A0-3-6 turnover procedure. This audit will include spot checks of critical system and parameter status during Plant operation. KM/sh/4 sala 8.
e Section 2.2.2.a - Control Room Access Trojan plant Administrative Order, A0-3-8, " Control Room Operations", has been revised and is in place in accordance with the PGE response to the NRC on October 17, 1979. The revised A0-3-8 explicitly states that the Control Operator has the authority and responsibility to restrict access to the control room. Also, the revised A0 indicates that trie Control Operator's decision of limiting access may be overruled by the Shif t Supervisor if necessary. The revised A0-3-8 also specifies personnel allowed access during emer-gency conditions to the Operations Supervisor, the onsite NRC and Scare resident inspectors, relieving shif t personnel and certain auxiliary operators. The line of succession for the person in charge of the con-trol room is specified and limited to persons possessing a current senior reactor operator's license. The lir.e of communication and authority for Plant management personnel not in direct command of operations is clearly defined in the chart of Trojan organization which is attached to the revised A0-3-8. KM/4sa10.lA9.
Section 2.2.2.b - Onsite Technical Support Center PCE will establish a permanent Technical Support Center (TSC) meeting the requirements of NUREG-0578 and clarifications specified in the NRC October 30, 1979 letter. Until the permanent TSC is established, an interim TSC will be utilized. A description of the interim and permanent TSC is detailed below. Interim 13C The Trojan Administration Building is designated as the interim TSC and is located onsite immediately adjacent to, but just outside, the security fence. The Administration Building is a well-engineered structure and will be habitable for a realistic spectrum of pctential nuclear plant accidents. The interim TSC will be activated at the Site Emergency accident classification in the current Trojan Radiological Emergency Response Plan (RERP). The control room and the Emergency Control Center (ECC) at the Visitors Information Center will serve as backup facilities in the event that the TSC has to be evacuated. Plans for staffing, per-sonnel assignments and evacuation of the TSC have been developed in the draft Revision 1 to the Trojan RERP submitted to the NRC on December 4, 1979. Since this plan is not finalized, interim procedures implementing the TSC have been incorporated into the Trojan plant procedures. Communications equipment installed at the interim TSC include in-Plant Executone, commercial telephone, and low-and high-band radio. A dedi-cated telephone circuit to the NRC is available in the Resident NRC Inspector's trailer adjacent to the Administration Building. The above communications equipment will allow the interim TSC to communciate with the control room, the Operational Support Center (OSC), the ECC, Westing-house and the NRC. Technical data including records that pertcin to the as-built conditions and layout of, structures, systems and components (FSAR, tech manuals, drawings, etc) are available at the interim TSC. Current Plant data will be verbally transmitted to the interim TSC by the Shift Technical Advisor stationed in the control room. Additionally, the closed-circuit televi-sion will be installed in the interim TSC by January 31, 1980 to monitor the control room. The interim TSC is equipped with portable radiation monitoring equipment for measuring both direct radiation and airborne radioactive contaminants. Action levels for evacuating the interim TSC are specified in the Trojan plant procecures. Permanent TSC The permanent TSC will be located in the control room visitors viewing gallery located directly above and adjacent to the control room. The viewing gallery is presently being used as a room where Plant visitors can observe control room instrumentation and activities. The viewing gallery is part of the Control Building and is a Seismic Category I structure. The viewing gallery has approximately 345 f t2 of floor -4 0 - 1703 274
space and 60 ft2 of counter top space which is sufficient to accom-modate 25 persons and information displays. Technical data (FSAR, tech manuals, drawings, etc) will either be stored in the viewing gallery or in a room in close proximity. The TSC will be equipped with in-Plant Executone, commercial telephone and dedicated NRC telephone. This equipment will provide the TSC with communication links to the control room, the OSC, the ECC, Westinghouse and the NRC Operations Center. Normal ventilation for the TSC is provided by the control room normal ventilation system. Emergency ventilation for the TSC will be provided by either modifying the control room emergency ventilation system or installation of an additional ventilation system. The control room emergency ventilation system is equipped with HEPA filters and charcoal adsorbers and is describ(d in Section 9.4 of the Trojan FSAR. Should a new ventilation system be installed for the TSC, that system will be equipped with HEPA filters, charcoal adsorbers and an emergency power supply. A new ventilation aystem would be designed to the same require-ments as the existing control room emergency ventilation system except that the system will not be redundant, Seismic I or instrumented in the, control room. The TSC emergency ventilation system will be automatically actuated following an Engineered Safety Feature Actuation System signal or high control room radiation alarm (>2.5 mr/hr). An area radiation and an airborne radioactivity monitor will be provided to monitor conditions within the TSC. These monitors will be provided with local readout and alarms. Plant procedures will specify the cri-teria for evacuation or potassium iodide administration for the TSC personnel in the event of high dose cates or high airborne radioactivity concentrations. PCE is following Westinghouse and NRC efforts in developing a list of parameters required to be displayed at the TSC. PGE will transmit to the NRC the list of parameters finally adopted industry wide. Prior to that time, PGE will transmit and display on one or more CRTs at the TSC those parameters considered appropriate by PCE using the NRC guidance in NUREG-0578. The current analog value or digital status of each parameter will be continuously available. Retrievable historical data for each parameter will be stored on magnetic tape. Data transmission equipment and power supply will be of high quality; comparable to control room equipment and power supply but not safety grade. Data transmission equipment interfacing P1nnt instrumentation will not cause degradation of control room or etaer Plant functions. TSC instrumentation power supply will be designed for continuous operation once the TSC is acti-vated. The data transmission system will be designed with provisioas for future transmission of the TSC data to offsite locations. Activation of the permanent TSC will occur at the Emergency Alert acci-dent classification level as described in Section 6.1 of the Trojan draf t RERP. TSC staf fing assignments and evacuation criteria are detailed in Section 5.2 of the Trojan draf t RERP. PGE will establish the permanent TSC prior to January 1,1981. MQH/4sa10A27 1703 275
Section 2.2.2.c - Onsite Operational Support Center As described in detail in the PCE response to the NRC dated October 17, 1979, PCE has established the onsite Operational Support Center (OSC) as of January 1, 1980. The Plant area designated as the OSC is the area behind the main control boards (Hagan Rack Area) on Elevation 93 ft in the Control Building, which is separate from but next to the control room with telephone communication lines and Exe:utone. The following personnel will report to the OSC: 1. Designated Operations personnel. 2. Designated Maintenance and Instrument Control personnel. 3. Designated Radiation Protection personnel. (Certain Radiation Protection personnel will report to the Access Control Point if it is habitable to ensure positive access control of potential high radiation areas within the Auxiliary and Fuel Building.) All other support personnel will report to the Visitors Information Center (VIC) in the event of a site evacuation. These support personnel may be dispatched from the VIC to the OSC if additional personnel are required in the Plant. The Trojan plant procedures have been revised to specify the existence of the center, and the methods and lines of communications in management. The Radiological Emergency Response Plan (RERP) will also be revised to specify the OSC when the current NRC review of the draft RERP is complete and finalized. 1703 276 KM/4sa10.lA8.}}