ML19199A562
| ML19199A562 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/18/1977 |
| From: | Jay Collins Office of Nuclear Reactor Regulation |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7905040143 | |
| Download: ML19199A562 (8) | |
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EC-22G HEPGRANDUM FOR:
S. Yarga, Chief, Lignt Water Reactors Sranen No. 4, OP" FRCit:
J. Collins, Chief, Effluent Treatment Systens Eranch, DSE SU3JZCT-ETSG TESTI!:0NY ON ISSUE NO. 21 huREG-0153, MIT:i RESPECT TO THE DESIGli GF THREE MIL: ISLAfD hJCLEAR ~ 'I:.1, UNIT NO. 2 In r2sponse to your memo of February 4,1977, sucject as above, I nava en. closed a copy of our draf t testim.ony concerninq :ne desicn of the radiation monitoring systens for the Three Mile Island. uclear Station, Unit "o. 2, wi'.a respect to Issue fio. 21 of NUREG-0153.
It should be notad that while ETSS has review resoonsioility for most of the topics covered by Issue No. 21, certain portions of the safety review fall witnin the review scope of the Electrical Instrumentation and Control Systeas Branca, Reactor Systems 3 ranch, Environmental Evaluation Branen, Containment Systems Branch, Accident Analysis Branch, and Radiological Assessment 3rancn. A thorougn discussion of all ascects of Issue 21 would require input or joint testimony fra all of tne branches noted above.
C T.C SU.C John T. Collins, Chief Effluent Treatment Systems Branch Division of Site Safety and Environmental Analysis
Enclosure:
Draf t Testimony 5eePage2 cc:
DISTRIBUTION:
DOCKET FILE NO. 50-320
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UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION THREE MILE ISLAND NUCLEAR STATION, UNIT N0. 2 DOCKET NO.
50-320 TESTI"0NY OF NRC STAFF ON ISSUE N0. 21 0F NUREG-0153 INSTRUME?tTS FOR MONITORING SOTH RADIATION AND PROCESS VARIABLES DURING ACCIDENTS WITH RESPECT TO THE DESIGN OF THREE MILE ISLAND NL' CLEAR STATION, UNIT NO. 2 BY PHILLIP G. STODDART EFFLUENT TREAT'4ENT SYSTE?iS BRANCH DIVISION OF SITE SAFETY AND ENVIRONMENTAL ANALYSIS OFFICE OF NUCLEAR REACTOR REGULATION 4
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INTRODUCTION This testimony is concerned with design aspects of radiation monitoring systems at Three Mile Island Nuclear Station, Unit No. 2, with regard to Issue 21 "Ir.struments for Monitoring Both Radiation and Process Variables During Accidents."
RESPONSE
Radiation monitoring systems, such as radioactive gaseous effluent monitors,. in-plant air::orne radioactivity monitors, and area radiation monitors, are not relied upon as primary means of coping with postu-lated high consequence accidents at Unit 2.
The detection, alam, and init4ation of necessary immediate actions for high consequence accidents are accomplished by measurement of other process variables, e.g., containme.: pressure.
Instrumentation system designs for these other process v.riables are reviewed by appropriate NRC staff to assure that initiation of all necessary actions, such as isolation of lines penetrating containment and initiation of emergency coolant flow, are perfomed automatically on tne basis of measurements of appropriate process variables.
The system designs are also reviewed to assure that the operator has available to him the minimum required infomation on the status of process variables to permit the assess-ment and control of the reactor and associated equipment in the post-accident period.
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Radiation monitoring systems have a number of functions at Unit 2.
These systems measure radiation levels and radioactivity concentra-tions during normal coerations, and anticipated operational occurrences, I
satisfying requirements for in-plant radiation protection, measurement of radioactive material in effluents, and environmental monitoring.
A number of these systems automatically initiate protective actions by other systems, e.g., actuation of ventilation filter systems to reduce releases of radioactive materials in gaseous effluents to values which are lower than the limits specified in the plant Techni-cal Specifications. This type of action comes under thr. category of anticipated operational occurrences rather than that af accident.
Lastly, the radiati a monitoring systems at Unit 2 provide radiological information to the operators to permit the diagnosing of plant radio-logical conditions folicwing an accident and to enable the coerators to determine the need for carrying out emergency procedures.
There are no radiation monitoring systems at Unit 2 that are required to automatically activate emergency equipment to mitigate the radio-At Unit 2, that logical consequences of the loss-of-coolant accident.
function is accomplished by the Safety Features Activation System (SFAS) upon detection of high pressure (4 psig) in the reactor building (containment).
The SFAS is a safety grade, fully redundant, Seismic Category I system and is reviewed by the staff under Standard Review Plan 7.3.
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@ At Unit 2, the reactor containment is nomally a suied system during cperation, with no inflow or outflow of ventilation air. At certain times, it is desirable or necessary to ventilate the containment while the reactor is in operation.
To accomplish this with minimal release of gaseous radiciodines and particulates, a recirculation system is actuated to cause the atmosphere in containment to flow through an atmosphere cleanup system consisting of high efficiency particul3te air (HEPA) filters and charcoal adsorbers. The HEPA filters remove radioactive particulates while the charcoal adsorcers remove radio-iodines. When the radioactivity concentration of containment atmosphere has been reduced to a level pemitting the air to be released without exceeding the limits specified in the plant Technical Specifications, valves are aligned to exhaust the containment air to the environment.
In this " purge" mode, the containment air passes again through filters and charcoal beds before being released through the plant vent. At the same time the purge is initiated, an air inlet valve is opened, allcwing fresh air to enter the containment. As the containment air is being purged, the exhaust stream is measured for radioactivity concentration in the purge duct downstream of the cleanup system.
It is again measured at a point downstream after mixing with the ventila-tion exhaust streams from other Unit 2 buildings. The exhaust air is then discharged through the plant vent.
In the event that measured radioactivity concentrations exceed predetermined values on either g
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monitor, an alarm from either monitor will cause the purge flow to be automatically diverted back to the recirculation or cleanup mode and automatically initiates the closure of all containment ventila-tion inlet ar.d outlet isolation valves.
The actions of the two monitors, i.e., the monitor on the containment ventilation purge exhaust duct and the monitor on the plant vent, serve to mitigate the consequences of radioactivity releases of a low order of magnitude and have the design function of preventing offsite gaseous radioactivity concentrations frc 1 exceeding limits specified in the plant Technical Specifications. Such releases are consifered to be in the category of " anticipated operational occurrences", as compared to normal operating conditions or accident condi tions.
In che staff's evaluation of the adequacy of gaseous effluent radio-activity monitoring system instrumentation with respect to the loss-of-coolant accident, the staff determined that gaseous releases from the reactor containment would be isolated from the environment by the action of the Safety Features Activation System and that there would be no flow of containment gases through either the reactor containment purge monitor or through the plant vent monitor. On tnis basis, there is no requirement for these monitors to have the extended range which would be required if reactor building gases were presumed to be released subsecuent to the accident.
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. reasons, it is not necessary that these monitors be redundant, be on emergency power supplies, or be designed as safety class equipment.
Therefore, we consider that the effluent radioactivity monitoring r.ystems provided for Unit 2, while adequate only for normal operating conditicns and for anticipated operational occurrences, need not be designed to meet special criteria for operation under accident condi-tions which are not applicable to Unit 2.
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