ML19210D361
| ML19210D361 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/20/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-30683, TAC-30684, TAC-30685, TAC-30686, TAC-30687, TAC-30688, TAC-30689, TAC-30690, TAC-30691, TAC-30692, NUDOCS 7911260351 | |
| Download: ML19210D361 (19) | |
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November 20, 1979 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 the supplemental infomation and justification for implementation schedules of the TMI Lessons Learned short-tem recom-mendations, which were requested in your letter dated October 30, 1979.
As stated in our response dated October 17, 1979, it is the intent of Portland General Electric Company to meet the requir ments and schedules set forth in the NRC letters dated September 13, 1979 and October 30, 1979. However, equipment implementation schedules largely depend on hardware availability, as well as the ability to effectively utilize scheduled Plant outages.
Some of the proposed modifications described in our letter dated October 17, 1979 require a Plant shutdown in order to perfom installation work.
The power supply situation in the Pacific Northwest is already critical due to low hydroelectric reservoir levels. Prospects for an improvement in the power supply situation are not expected for some time. The only alternative energy is by oil-fired generation, and a small amount of natural gas, with limited availability and at a very high cost.
It is thus our firm belief that a Plant shutdown is not in the public interest and t. hat the interim solutions presented in our October 17 letter, together with the supplemental infomation presented in the attachment, provide sufficient justification for the course of our action and reason-able assurance for safe operation during the interim period.
The attachment delineates the supplemental infomation and justification of the implementation schedules committed for the short-tem actions, ON 17QI dbJ7 Slll
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Mr. Harold R. Denton November 20, 1979 Page two which proposed either alternative completion dates or a different method of implementation than the NRC guidelines. The implementation schedules of the remaining short-term actions are in compliance with NRC require-ments, and the descriptions of these items were provided in our response dated October 17, 1979.
In addition, the attachment also contains a supplemental description of Shift Technical Advisors in accordance with the commitment in our letter dated October 17, 1979 and additional information requested by your staff during the telephone conference on November 14, 1979.
Sincerely, Af C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation and
- ~aintenance CG/KM/mg/4kkSA21 Attachment c:
Mr. Lynn 7 rank, Director State of Oregon Department of Energy Mr. A. Schwencer, Ct.ief Operating Reactors Branch #1 Division of Operating Reactors U.S. Nuclear Regulatory Commission Mr. R. H. Engelken, Director U.S. Nuclear Regulatory Commission Region V jzq
e ATTACHMENT TROJAN NUCLEAR PLANT Supplemental Information and Justification for Implementation Schedules on TMI Short-Term Actions 2.1.3.a Direct Indication of PORV and Safety Valve Position The PORV limit switches will be upgraded : or post-LOCA environmental conditions (in accordance with Trojan FSAR), and new valve position indicating devices will be installed on safety valves with the quali-fication requirements described in our response (October 17, 1979).
The above installation work will commence during the first cold shutdown after December 1, 1979 and will be completed no later than the startup of Cycle 3 operation in 1980. Any necessary additional qualification tests and documentation will be completed by June 1, 1980 The current schedule for equipment delivery for safety valve position indication (acoustic sensors) is December 10, 1979. A construction time of approximately 32 man-weeks is estimated. A Plant shutdown is required for this installation work.
Because of the critical power situations and material availability, as mentioned in the cover letter, the commitment date is considered to be the earliest practical date for completion.
With regard to Item 2 in the CLARIFICATION SECTION of the hRC letter dated October 30, 1979, the following is supplemental information to our response dated October 17, 1979:
2.
Valve positions for both PORVs and safety valves will be indicated in the control room.
An alarm will be provided in conjunction with the indication for the safety valves; however, there are presently no plans for adding an alarm in addition to the existing indication for the PORVs.
Unlike tba safety valves, the PORVs are operable from the control room and may be used by the operator under certain postulated conditions for primary system control. The unsafe (alarm) position could be either open or closed dependent upon Plant conditions. We do not consider an alarm to be either desirable or necessary for the PORVs; however, we will reconsider this requirement if further clarification could be offered on what should be alarmed and when.
As stated in our response (October 17, 1979), the PORVs currently have seismically qualified stem position indicators which provide reliable direct position indication in the control room. Although safety valves are not equipped with positive valve position indicators, the following present design features provide indirect position indications for safety valves as well as PORVs:
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265 1.
Temperature is sensed in the discharge manifold for the two PORVs and in the discharge pipe of each of the three safety valves, and is displayed and alarmed in the control room (see FSAR Figure 5.1-1).
2.
The pressurizer relief tank has sensors for high pressure, high temperature and high water level with indication and alarms in control room.
Existing monitors for these parameters, together with increased operational attention as an outcome of the TMI accident, ensure safe Plant operation until modifications are completed.
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2.1.3.b Instrumentation for Detection of Inadequate Core Cooling The emergency procedure (EI-1, " Loss of Reactor Coolant") has been revised and presently requires operators to maintain subcooling margin based on manual calculation using the installed pressure and temperature instruments. The pressure-temperature curve which incorporates a mar-gin to saturation condition is attached to the emergency procedure for operator usage. This method is sufficient to ensure adequate subcooling condition in the Reactor Coolant System during the interim period.
As stated in our October 17, 1979 response, it is our intent to provide reactor coolant Subcooling Margin Meters (SMMs) in accordance with NUREG-0578 requirements and the guidelines in the October 30, 1979 NRC letter. A Plant shutdown will be required to implement this modification; certain work has to be done in the control room and in the Containment during shutdown. The purchase order of the SMMs has been issued and a verbal commitment of the equipment delivery date by the manufacturer is December 17, 1979. The required time for development of final design, installation, system testing and development of the required operating procedures is 6 weeks after the equipment delivery. The installation of the SMMs will be completed by the startup of Cycle 3 in 1980. Consider-ing the above and the critical power situation in the Northwest, the startup of Cycle 3 is the earliest feasible completion date for this system.
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267 2.1.4 Containment Isolation For the modification of 23 Containment isolation valves, the current best estimate of material delivery is at the beginning of February 1980 and the construction time is approximately 42 man-weeks.
Some of the circuit modifications, as described in our response dated October 17, 1979, must be performed inside Containment (preferably during cold shutdown); they cannot be made with the Plant at power. Because of the material avail-ability and the same reasons mentioned in Section 2.1.3.b, the earliest possible date for completion of this modification is by the startup of Cycle 3, 1980.
In the interim period, administrative controls have been implemented to require operators to make positive verification that the 23 affected valves are in the closed position prior to resetting a Containment isolation signal. This administrative control will ensure Containment integrity until modifications are complete.
b0 2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactivity The commitment date of January 1, 1980 for development of the leak reduction program is consistent with the NRC position. The description of this program, including the systems to be inspected, was previously provided in our response to the NRC dated October 17, 1979. Estimates of current system leak rates will be provided with the January 1, 1980 submittal.
Testing using the new leakage reduction program will be performed at the first opportunity af ter January 1,1980 but, in any event, no later than the startup of Cycle 3 in 1980. This schedule is based on the fact that the tests for some systems may have to be conducted during Plant shutdown or during certain system conditions which are not possible to schedule precisely at this time.
17oi 269 2.1.6.b Plant Shielding Review The design review of the radiation shielding will be conducted in accor-dance with NRC requirements as stated in our response to the NRC on October 17, 1979.
It is expected that the design review will be completed by January 1, 1980. We will attempt to complete any Plant modifications or changes in operating procedures resulting from the design review by January 1, 1981; however, the extent of structural modifications or changes in equipment arrangement will not be known until the shielding analyses are completed. An additional schedule for modification work will be provided to the NRC by January 1, 1980.
2/0 2.1.8.a Improved Post-Accident Sampling Capability The PGE response dated October 17, 1979 provided a description of imple-mentation method and schedule of this action, which are in accordance with the NRC requirements dated September 13, 1979.
Based on the clarifications provided in the NRC letter dated Occober 17, 1979, the onsite radiological and chemical analysis facility will include the capability to provide quantification of dissolved gases in the liquid.
Hydrogen levels in the Containment atmosphere are analyzed by the redun-dant, Seismic Category I post-accident hydrogen analysis system discussed in Section 2.1.9 of our response.
1Toi 271 2.1.8.b Increased Range of Radiation Monitors On October 17, 1979, PGE respondea to this NRC position by noting that high-range monitors (300 pCi/cc Xe-133) were already scheduled for installation on the Auxiliary Building and Condenser Air Ejector Exhausts. The Containment radiation monitoring system has an existing upper range of 300 yCi/cc. The October 17, 1979 response included the bases for concluding that 300 pCi/cc is en acceptable upper range for these radiation monitors, based on specific Trojan characteristics.
PCE has reviewed the clarification provided in the NRC October 30, 1979 letter from Harold R. Denton to all operating nuclear power plants and has concluded that an upper range of 300 pCi/cc is acceptable. This con-clusion is based on the following information which is in addition to the information provided on October 17, 1979:
A.
Auxiliary Building:
1.
Using FSAR ventilation flow rates, a concentration of 250 pCi/cc in the Auxiliary Building exhaust cor-responds to the specified release rate of 10,000 Ci/sec. This concentration is below the upper range of 300 pCi/cc.
2.
The 300 uCi/cc upper range is more than two orders of magnitude above the actual release concentrations for the Three Mile Island incident (1 pCi/cc) reported in NUREG-0578.
3.
A concentration of 300 pCi/cc in the Auxiliary Building ventilation exhaust corresponds to a site boundary gamma dose rate of approximately 250 rem /hr. Thus, the 300 pCi/cc upper range is well above the level required to identify signficant offsite consequences of an accident and to evaluate offsite doses versus the EPA Protective Action Guides.
4.
PGE previously committed to providing a flow path from the waste gas decay tank discharge into the Containment.
Therefore, in the event of an accident, high level gase-ous activity vill be transferred to the Containment and not discharged via the Auxiliary Building ventilation exhaust.
In addition, the 300 uci/cc upper range monitor will remain on scale in the event of a waste gas decay tank rupture even if the tank contained the maximum radioactivity permitted by the Environmental Technical Specifications.
B.
Containment :
1.
There are three functions which can be postulated for the Containment Radiation Monitoring System. These functions include purge exhaust monitoring, internal Containment monitoring and Hydrogen Purge System 'to1 2/2
monitoring. The October 17, 1979 submittal addressed the acceptability of the present system for Hydrogen Purge System monitoring.
2.
The Containment Purge Exhaust System will be isolated in the event of an accident. Following the accident, the Containment purging will not be performed until airborne radioactivity levels inside the Containment have decayed well below 300 pCi/ce. Therefore, the present monitor range is acceptable for this function.
3.
The non-Seismic Category I Containment Radiation Moni-toring System will not be used to monitor airborne radioactivity levels inside the Containment following a large accident (airborne levels above 300 pCi/cc) because of the potential of violating Containment integrity.
Airborne radioactivity levels inside the Containment will be measured using the two redundant area radiation monitors which are to be installed prior to January 1, 1981. Conversion factors will be pro-vided which will convert the area radiation monitor readout to radioactivity concentrations. The capa-bility will exist to measure 105 pCi/cc (Xe-133) within the Containment in compliance with the NRC position.
C.
Condenser Air Ejector Exhaust:
1.
The Trojan Technical Specifications limit the failed fuel rate to approximately 0.27 percent.
Calcula-tions using the assumptions of the Trojan FSAR and this failed fuel rate demonstrate that simultaneous ruptures of at least two steam generator tubes could be postulated without exceeding the upper range of this monitor.
2.
In the event of accidents that exceed the Plant design bases, such as a steam generator tube rupture coinci-dent with core damage, the operator action would be to isolate the affected generator and to depressurize the Reactor Coolant System to below the relief and safety valve setpoints of the steam generators, thereby ter-minating the release.
Further, it is most likely that any steam generators with ruptured tubes would be isolated soon after the accident initiating the event, and well before the occurrence of any core dauage.
In summary, PGE considers that the 300 pC1/cc upper range for these three monitoring systems, backed up by the capability to determine higher levels through effluent grab samples, provides an adequate capability to measure post-accident effluent releases.
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For the Containment area radiation monitors, current estimate of the equipment delivery date is August 1980, and installation work and system testing will require 4-6 weeks. A Plant shutdown will be required to install this system. We intend to complete the installation work by January 1, 1981; however, should the power supply situation become criti-cal again in 1980, the installation work will commence at the first cold shutdown af ter September 1,1980 (subject to equipment availability) and will be completed no later than the startup of Cycle 4, currently sched-uled for the summer of 1981 (reference October 17, 1979 PGE response).
The interim procedures for qualifying accidental radioactivity releases will be implemented in accordance with the NRC requirements and schedules.
'7')1 274 2.1.8.c Improved In-Plant Iodine Instrumentation As described in our response dated October 17, 1979, the mobile low bcckground counting facility equipped with a Ge-Li gamma spectroscopy system for counting high activity sources will be established onsite, but outside Plant structures, by January 1, 1981.
As an interim measure, until the mobile counting facility is established, a single channel analyzer will be provided for use in the Plant prior to January 1, 1980 to promptly analyze airborne samples for iodine.
f5 2.1.9 Analysis of Design and Off-Normal Transients and Accidents The implementation schedules in our response dated October 17, 1979 for necessary procedures and operator training were in accordance with the schedules established by the NRC.
The description of instruments for Containment pressure and water level was provided in our response dated October 17, 1979.
The current esti-mate of equipment delivery dates is August 1980 for bot.: Containme nt pressure and Containment water level. The construction work for those instruments is estimated to be 4-6 weeks.
"e are attempting to complete all installation work by January 1,1981; however, should the power situation remain critical throughout 1980, the installation work will commence at the first cold shutdown after September 1, 1980 (subject to the equipment delivery) and will be completed no later than the startup of Cycle 4 in 1981 (reference October 17, 1979 PGE response).
For the Containment water level indication, further investigation of the means to provide the water level indication has led to a different design approach. We are now considering a displacement transmitter as the detection instrument. This instrument is currently used elsewhere in the Plant and is providing reliable service. As before, narrow-range indication will be provided to monitor water level in the Containment sumps. Wide-range indication will be provided to monitor water level f rom the top of the sump to the top of the flood level (approximately Elevation 53 ft).
Narrow range and wide range will be combined into one detector and one indicator if the instrument being investigated has the required characteristics. The indicators are to be mounted on new panels to be located on the north wall of the control room.
Design of the Containment instrumentation for pressure and water level will be in accordance with the Trojan licensing commitment in the FSAR.
Additional qualifications will be incorporated into design as much as practical, if they are available.
During the interim period, the Trojan plant currently has four safety-grade pressure transmitters for measuring Containment pressure as high as 75 psig, which is above the design pressure for the anticipated accident inside the Containment. For the Containment water level indication, each of the two Containment sumps has a water level switch with a measurement range from the bottom to the top of the sump.
b Yb 2.2.1.a Shif t Supervisor's Responsibilities The first Management Directive and review of administrative duties of the Shif t Supervisor will be completed by the Assistant Vice President, Ther-mal Plant Operations and Maintenance on or before January 1,1980. The Administrative Orders will be modified and the indoctrination of the Shif t Supervisor will be completed by January 1,1980.
In the revised Administrative Order, the Shift Supervisor will remain in the control room during accident situations until the RCS conditions are stabilized or he can be properly relieved.
'?o1 277 2.2.1.b Shift Technical Advisor As committed in our response dated October 17, 1979, the following is the detailed description of the Shif t Technical Advisor (STA) program.
In reply to NRC short-term Three Mile Island recommendations, we will assign a STA to provide support to the operating crews in the event of an accident and to perform the operations assessment function. Our commit-ment is to develop and implement an initial program by January 1,1980 and then to expand this program by January 1, 1981 to include in-depth training of the personnel performing the Shift Technical Advisor function.
We have developed the following two phase plan to implement this program:
Phase I - Short-Term Plan In order to meet the January 1, 1980 date, we will use person-nel already employed in the Plant Engineering Staff and other engineering positions at the Trojan plant. Our initial program will establish a pool of Plant Engineering personnel. Each day, one of the pool members would be assigned to function as the STA. This would be a 24-hr assignment and the engineer assigned would be onsite for this perfid. Arrangements for sleeping and eating facilities will be cade and the individual will be allowed to sleep during an 8-hr portion of his onsite pe riod. All of the engineers in the pool will be rotated on a day-to-day basis through this assignment and will be able to scapond upon notification from the control room. The STA will be physically present and able to assess the condition of the Plant within the 10-min response criteria.
Experience with off-hours notification of management and technical people has shown that sufficient mental alertness can be achieved within 10 min to be completely functional.
An alarm notification system, such as a portable pager, or
" beeper", will be used to summon the STA to the Control Room.
The operating crew will actuate the notificat,n system when the STA's presence is required.
Phase II In the longer term, individuals performing the STA function must be knowledgeable in those academic and operational disci-plines essential to plant safety. To this end, a candidate's experience and education will be considered in the development of the training program. A fully qualified STA will have training in the following areas:
Fundamentals of thermodynamics, heat transfer and fluid flow Fundamentals of electrical sy.tems 1?Qi
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- Reactor physics and kinetics
- Reactor core, and internals design, including vessel design
- Plant transient and accident analyses, including observation of transients and accidents at a suitable simulator Knowledge of Plant system design and principles of operation, including instrumentation and protection systems We estimate the time interval to complete this initial training for a typical degreed member of the technical staff to be about 2 months.
In the future, the SIA function may be performed by non-degreed, but appropriately qualified (SRO) individuals.
Duties and Responsibilities The STA will provide the accident assessment function as well as the operational assessment function.
In the accident assessment role, he will be available within 10 min to evaluate off-normal events during the course of an unusual transient or accident. He will be available to advise the Shif t Supervisor, but will assume no active role in the direction of operator activities while performing his duties.
In the operational assessment role, he will, during periods of normal Plant operations, perform duties that will enhance the safety and reliability of the Plant. Typical assignments to the STA to accomplish these duties may include:
Review of Licensee Event Reports of other nuclear facilities
- Review surveillance test procedures and data
- Review proposed Plant design changes from the viewpoint of operational safety Results of these reviews will be forwarded to appropriate Plant supervisors for resolution or action. Organizationally, the STA will be independent from the operations group and report to the Plant Engineering Supervisor, who in turn reports to the Manager, Technical Services.
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2.2.1.c Shift and Relief Turnover Procedures As discussed in our response dated October 17, 1979, maintenance and testing of equipment that by themselves could degrade a system critical to the prevention and mitigation of operational transient and accidents or initiate an operational transient is currently co strolled by a main-tenance request.
In order to ensure proper turnover of maintenance and testing in progress at shift change, auxiliary operators and technicians providing shif t coverage will maintain a log of critical equipment and components under maintenance and test.
'7'i 280 2.2.2.b Onsite Technical Support Center A preliminary design review of the Technical Support Center has been initiated, and it is intended that PGE will satisfy the NRC requirements and the implementation schedules; however, since the design criteria and the design of the facility is presently in an early stage, it may not be possible to complete all the work by January 1,1981. A detailed sched-ule for implementation will be provided by March 1, 1980.
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