ML19351E471

Supplemental Testimony in Response to ASLB Question 6 Re Emergency Feedwater Reliability.Prof Qualifications & Certificate of Svc Encl

ML19351E471
Person / Time
Site:	Three Mile Island
Issue date:	11/25/1980
From:	Colitz J, Keaten R, Ross M GENERAL PUBLIC UTILITIES CORP., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8012100177
Download: ML19351E471 (25)

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD *

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3 In the Matter of )

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METROPOLITAN EDISON COMPANY ) Docket No. 50-289

) (Restart)

(Three Mile Island Nuclear )

Station, Unit No. 1) )

LICENSEE'S SUPPLEMENTAL TESTIMONY OF ROBERT W. KEATEN, JOSEPH J. COLITZ AND MICHAEL J. ROSS IN RESPONSE TO BOARD QUESTION NO. 6 (EMERGENCY FZ2DWATER RELIABILITY) ,

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, OUTLINE.

This testimony supplements Licensee's Testimony of Gary R. Capodanno, Louis C. Lanese and Joseph A. Torcivia in Response to Board Questions 6.a, 6.b, 6.c, 6.g, 6.h, 6.i, 6.j and 6.k dated October 21, 1980 and Licensee's Testimony of Robert.C. Jones, Jr. ir Response to Board Questions 6.e and 6.f, dated October 28, 1980. In particular, this testimony is in response to the Board's clarification of Board Question C and addresses the means'by which the emergency feedwater system brings the plant to cold shutdown, the complexities and problems involved in the operation and termination of the feed and bleed cooling mode,.and initiation of an alternative cooling mode to the feed and bleed mode.

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- DESCRIPTION OF CORE COOLING AND HEAT REMOVAL PROCESSES . . . . . . . . . . . 2

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METHODS'OF ACHIEVING COLD SHUTDOWN- . . . . . .. . . . . . . . . 8 OPERATION AND TERMINATION I OF FEED AND BLEED COOLING . . . . . . . . . . 10 RESTORATION OF EMERGENCY FEEDWATER. . . . . . . . . . . . . . . .

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INTRODUCTION Licensee's initial response to Board Question No. 6, which addresses emergency feedwater reliability, was presented in " Licensee's Testimony of Gary R. Capodanno, Louis C. Lanese and Joseph A. Torcivia in Response to Board Questions 6.a, 6.b, 6.c, 6 g, 6.h, 6.1, 6.j and 6.k," dated October 21, 1980,

" Licensee's Testimony of Robert C. Jones, Jr. in Response to Board Questions 6.e and 6.f," dated October 28, 1980, and "TMI-l Emergency Feedwater System," Licensee's Exhibit No. 15.

The Board, at-the hearing session of November 5, 1980, cJarified the issues which it intended to be addressed in Board Question No. 6 to include the following:

How would the emergency feedwater system, if relied upon, bring the plant to cold shutdown?

If emergency feedwater fails. what are

the complexities and problems involved in the operation and termination of the feed and bleed cooling mode?

How is an alternative cooling mode, such as restoration of emergency feedwater, initiated in order to bring the plant to cold shutdown?

See Tr. 4812, 4813.

This testimony, by Mr. Robert W. Keaten, GPU Manager of Systems Engineering, Mr. Joseph J. Colitz, TMI-l Manager of Plant Engineering, and.Mr. Michael J. Ross, TMI-l Supervisor of Operations, is addressed to the Board's inquiry in Board Question No. 6.

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SUPPLEMENTAL RESPONSE TO BOARD QUESTION NO. 6 BY WITNESSES KEATEN, COLITZ AND ROSS:

Earlier witnesses on behalf of Licensee have discussed the various processes available for core cooling and removing residual heat from the primary coolant at TMI-1. In order to summarize these processes, and to assist in responding to the Board's inquiry, a diagram is attached (Figure 1) which illustrates the core cooling and heat removal processes. A simplified schematic drawing of the plant which illustrates the key features of these processes is also attached (Figure 2) .

Description of Core Cooling and Heat Removal Processes The fuel, in the reactor pressure vessel, is contained in a closed system of circulating water known as the primary or reactor coolant system (RCS). The reactor coolant normally removes heat from the fuel and transports it through two piping loops (hot legs) to the top of the two steam generators (also called, Once Through Steam Generators (OTSGs) ); the cooler fluid then goes out the steam generator cold legs, through four reactor coolant pumps, and back into the reactor vessel and the lower portion of the core. The nominal capacity of either steam generator to remove heat is 50% rated reactor power and is, therefore, more than adequate to remove all residual-heat. The reactor coolant system can transfer residual heat to the steam generators with or without reactor coolant pumps operating.

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.The two steam generators are large, vertical, tube-in-shell heat exchangers that transfer the primary system heat through tubing walls into a secondary system. The primary coolant, normally in liquid form, passes through the inside of the steam generator tubes. Heat is transferred through the tube surface to the outer, or secondary, side of the tubes where the cooler secondary fluid is heated. The secondary coolant boils in the

steam generators. Secondary side makeup water (feedwater) is normally provided by the main feedwater system. The feedwater system contains two main feedwater pumps, three condensate pumps and three condensate booster pumps located in the turbine building which supply the two steam generators. This system can supply enough feedwater to remove residual heat with only one main feed-water, one condensate pump and one condensate booster pump supplying one steam generator. The steam produced in the steam generators is normally piped through the containment structure and through the turbine bypass valves to the shell side of a condenser where it is condensed to liquid water. From there the water is returned to the steam generator by the main feedwater system.

Cooling for the condenser is supplied by a circulating water loop, which finally discharges heat to che atmosphere via the natural draft cooling towers.

The Emergency "eedwater (EFW) system at TMI-l is an alternate source of steam generator secondary side water supply, In the event main feedwater is not available (e.g., the proper combination of the condensate pumps, condensate booster pumps, r,

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main feedwater pumps, or the main condenser are not available), the EFW system would supply water from either or both of the condensate storage tanks to the secondary side of the steam generators. The steam produced would be removed through the turbine bypass valves to the main condenser, if available, or through the main steam relief valves or the atmospheric dump valves to the atmosphere.

The two motor driven emergency feedwater pumps can be powered from either on-site or off-site AC power sources. The steam driven emergency feedwater pump requires neither off-site nor on-site AC power sources to operate. Any one of the three EFW pumps supplying water to either of the two steam generators has sufficient capacity to remove residual heat.

The primary system normally operates at a pressure above that at which boiling occurs;.i.e., the coolant is subcooled, or below the saturation temperature. A pressurizer, which contains a cushion of steam, is attached to the primary system to maintain pressure within normal operational limits by heating its volume of water with electric heaters (pressurizer heaters) or by cooling the steam region with a water spray (pressurizer spray) .

Two code safety valves, located at the top of the pressurizer, are designed to open (automatically) without external signals or power and to release steam when primary system pressure approaches normal design operational limits. In addition, a power-operated relief valve (PORV) is present to open prior to the code safety valves, thus minimizing the frequency of operation of the code safety valves.

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Three high pressure injection pumps located in the Auxiliary Building are provided to add inventory via the RCS cold legs to the primary system at high pressure, one pump normally operates to replenish water which is continually being removed from the reactor coolant system for purification, chemistry control and by reantor coolant pump seal leakage.

At low reactor coolant system pressures and temperatures during normal shutdowns, residual heat is removed by the decay heat removal system. Low pressure injection (LPI) pumps (also called Decay Heat Pumps) are used to provide closed loop cooling by circulating primary coolant through a heat exchanger. The residual heat from the LPI coolant loops is transferred to the river via a second system, the Decay Heat Closed Cooling Water system. There are two independent heat removal trains as described, each capable of removing all residual heat. The pumps in these trains are operable from either off-site or on-site power.

In the event of a loss-of-coolant accident, the emergency core cooling system (ECCS) cools the core by replenishing reactor coolant inventory. The ECCS includes the high pressure injection (HPI) and low pressure injection (LPI) systems, and two core flood tanks (CFTs). Under accident conditions when the HPI system is called upon to operate, it injects water, taken from the Borated Water Storage Tank (BWST), into the reactor coolant system at high pressure. The CFTs and LPI system inject water at lower system pressures. The CFTs are pressurized with nitrogen, require no power to function and automatically inject a limited volume of water into the primary system when the primary system pressure drops below 600 psia. The LPI system can operate in two modes. It can pump water into the reactor pressure vessel, in a manner similar to HPI operation, from the BWST and in the longer term from the reactor (containment) building sump. It can also feed the HPI pumps from the sump, if RCS pressure remains above the capability of the LPI pump. Following a loss-of-coolant accident, the prinary coolant which collects in the reactor building is cooled oy the decay heat system heat exchangers before being reinjected into the reactor coolant system by the LPI or HPI pumps.

In the case of a normal reactor trip, the process of removing the decay or residual heat from the primary or reactor coolant system would be through the steam generators to secondary coolant provided by either of the feedwater supply systems. Assuming an end of life, equilibrium full power history before the time of trip, the decay heat level is approximately 7% of full power at the time of trip. This heat level quickly decays to 4% within 40 seconds and roughly to 1% in an hour.

An equivalent percentage of main feedwater flow would be required to maintain equilibrium RCS temperature, or approximately 720 gpm of emergency feedwater 40 seconds after trip. The flow requirements and capabilities of the main feedwater pumps are above 50% of full rated power. Consequently, there is abundant capacity in either of the two main feedwater pumps to provide feedwater flow for residual heat removal.

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O If main feedwater is unavailable, the EFW system will provide sufficient secondary coolant. The EFW system has two flow paths, supp1.ied by one turbine-driven pump and two motor-driven pumps, which can supply emergency feedwater to either or both of^the steam generators. (See Licensee's Exhibit No. 15 for a complete description of the TMI-l EFW system.) The turbine-driven pump has a rated capacity of 920 gpm, and each motor-driven pump has a rated capacity of 460 gpm. Either one turbine-driven or both motor-driven pumps exceed the requirements to remove the 7% residual heat that exists at the time or reactor trip. By 2 1/2 minutes after trip, one motor-driven pump has enough capacity to remove the decay heat. Even if only one motor-driven pump were available initially, adequate heat removal would be provided. RCS temperature and pressure would initially increase, possibly resulting in lifting a relief valve. As decay heat drops, the EFW pump would supply enough water to overcome the temperature / pressure rise and restore normal conditions.

The TMI-l EFW system at restart will have redundancy, diversity and sufficient capacity to act as a water supply for reacto. acolant system cooling under the normal single-failure assumptions applied to safety-grade systems. (See Licensee's Testimony of Gary R..Capodanno, Louis C. Lanese and Joseph A.

Torcivia in Response to Board Questions 6.a, etc., October 21, 1980.)

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Finally, as discussed in Licensee's testimony in response to UCS Contentions 1 and 2, even if no feedwater is available (i.e., all main feedwater and all emergency feedwater flow has been lost) the core can be adequately cooled simply by maintaining a sufficient inventory of water in the' reactor vessel. This is accomplished by using the high pressure injection pumps to feed water from the Borated Water Storage Tank into the reactor coolant system, so that the core is covered with water or a two-phase mixture of water and steam. If no feedwater is available, the reactor coolant system pressure will increase to the setpoint of the relief valves, at which point one or more relief valves will open to control the pressure. This combination of use of the high pressure injection system to maintain adequate water inventory and use of relief valves to control system pressure is referred to as feed and bleed cooling.

Methods of Achieving Cold Shutdown The above processes basically describe the methods available for decay heat removal immediately following reactor trip while the system is still at or near normal system temperature and pressure. Several methods are available to proceed to cold shutdown from_this condition depending on the remaining operable equipment. However, it should be noted that the plant can remain in the hot condition for extended periods with any of these methods is the decision to transition to cold shutdown is deferred.

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The normal method for cooldown from operating pressure and temperature is to remove steam from the steam generators at-a rate greater than decay heat, using the main feedwater system, the turbine bypass valves, and the main condenser. This is accomplished by taking manual control of the turbine bypass valves and opening the valves to a position where the resulting steam flow to the condenser yields the desired cooldown rate of the reactor coolant system. This method can be maintained despite single active failures in the process train ,inclraing single failures in offsite power feeds. The reactor coolant system can be cooled by this method to the point that the decay heat removal system is put into operation (about 250'F/320 psig).

The decay heat removal system can then continue the normal shutdown cooling process until the conditions of cold shutdown are reached (Tave (200*F).

If the main feedwater system is lost, the Emergency Feed-water System can provide the same capability to ultimately cool down the reactor coolant system. If the condenser is available, the secondary system will function as a closed loop by steaming through the turbine bypass valves to the condenser and water drawn from the condenser by the emergency feedwater pumps and returned to the steam generators. If the condenser is not available, steam can be released to the atmosphere via the atmospheric dump valves. These valves can be controlled in the same manner described above for the turbine bypass valves in order to achieve the desired ccoidown rate. In this cooling mode water from the condensate storage tanks is fed to the steam generators by the emergency :.3edwater systems and then released to the atmosphere, The condensate storage tanks are required by the technical specifications to have 150,000 gallons in each tank during reactor operation. This amount of water is more than adequate to allow the reactor coolant system to be cooled to the temperature and pressure where the decay heat removal system can be placed in operation, prior to the depletion of inventory in the condensate storage tanks.

Operation and Termination of Feed and Bleed Cooling

Initiation of the feed and bleed cooling mode is a very simple operation. If neither main nor emergency feedwater is available, the operator will initiate and maintain full high pressure injection until feedwater is restored. He can open the RC-RV-2 (PORV) and RC-V2 or allow the code safety valves to open to provide a flow path.

Once initiated, the feed and bleed cooling mode will automatically continue without need for additional short term I

operator actions. In the long term the operator must transfer the suction of the high pressure injection pumps from the BWST

to the containment building sump via the low pressure injection I pumps. If ESFAS has automatically initiated, this transfer requires opening 4 valves and closing 4 valves all of which can be done at the main control console. If ESFAS has not automatically initiated, the LPI pumps must be started manually-but this also can be accomplished from the main control console.

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Termination of the feed and bleed cooling mode is also very simple.' Once the appropriate criteria are met the HPI discharge valves are throttled and eventually the HPI pumps are turned off. These actions are also performed from the main control console. Such throttling and/or termination of high pressure injection, however, is only permissible when specific criteria regarding RCS conditions are met. (See Licensee's testimony in response to UCS Contention 10.)

It should be noted that the s!mple actions associated with

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initiation, continuation and termination of feed and bleed cooling would be performed by an operator assigned to this portion of the control panel. Any parallel actions being taken in an attempt to restore main or emergency feedwater would be taken by a different operator assigned to the feedwater control panel. The TMI-l technical specifications require that two licensed reactor operators be in the control room at all times the plant is in operation.

The normal control room practice is that immediately upon reactor trip one operator goes to the portion of the console from which HPI and LPI are controlled, and the other operator goes to the feedwater control portion of the panel. This allows actions to be carried out in parallel under the supervision of the senior watchstanders.

Restoration of Emergency Feedwater If no feedwater is available, and the plant is operating in the feed and bleed mode, the normal steps taken would be directed at restoring emergency feedwater flow, as described in l

the follow-up action section of EP 1202-26A. The exact steps depend upon the reason why no feedwater is available and generally consist of verifying that valves are in the correct position, verifying that the pumps have started and taking manual actions where pump or valve actuation have not occurred correctly.

Assuming emergency feedwater is made available, the steam generator can be restored as a heat sink by adding emergency feedwater to the steam generator (s) and relieving steam through one or both atmospheric dump valves or through turbine bypass 4

valves to the condenser. These pumps and valves are normally ,

operated from the control room but the valves can also be operated locally and the steam-driven escrgency feedwater pumps can be started locally. With the steam generator in operation, primary system temperature can be reduced below system saturation temperature and a 50' subcooling margin will be maintained or reestablished. HPI can then be throttled, and a bubble can be formed in the pressurizer by energizing pressurizer heaters and reducing high pressure injection flow to allow the PORV or primary safety valve (s) to close. The normal makeup system can be used.

Once the bubble has been reformed in the pressurizer, the plant has been returned to a normal shutdown condition and cooldown may continue using normal plant cooldown procedures.

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l JOSEPH J. COLITZ l Business Address: Metropolitan Edison Company Three Mile Island Nuclear Station P.O. Box 480 Middletown, Pennsylvania 17057 Education: B.S., Mechanical Engineering, Villanova University, 1963 Post-graduate courses in Reactor Engineering and Health Physics, University of Michigan Exnerience: Manager - Plant Engineering, TMI-1, Metropolitan L;ison Company, 1979 to present. Responsible for providing technical engineering support for all aspects of TMI-l operations via review and evaluation of changes to procedures, systems and equipment and their relationship to licensing design basis criteria.

Director of Projects, Generation Department, Metropolitan Edison Company, Reading, Pennsylvania, 1977 to 1979. Responsibilities included industrial waste plants at company fossil units, the backfit of a cooling tower to a fossil unit and the installation of the TMI security system. Following the TMI-2 accident, served as the back-shift senior on-site representative for TMI-2 activities.

TMI-l Unit Superintendent, Metropolitan Edison Company, 1974 to 1977. Responsible for the overall operation and maintenance of TMI-1, including plant engineering and health physics.

Licensed as a Senior Reactor Operator on TMI-1.

Plant Engineer, TMI-1, Metropolitan Edison Company, 1973 to 1974. Responsible for all mechancial, electrical, nuclear and instrumenta-tion and control engineering for TMI-1.

Supervisor of Operations, Three Mile Island Nuclear Station, Metropolitan Edison Company, 1968 to 1973. Involved with the initial selection and training of operating personnel, preparation of plant operating procedures and support for the startup and test program.

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Engineer, Metropolitan Edison Company, 1967 to 1968. ' Spent 1 1/4 years at-the Saxton Nuclear Station.in training on the operation and maintenance of a nuclear station. Licensed by the NRC as a Reactor Operator.

Cadet Engineer and Engineer, Metropolitan Edison Company, Reading,. Pennsylvania, 1963 to 1967. Served in various positions' relating to fossil plant engineering, including Plant Engineer of the Crawford Generating' Station.

Professional Affiliations: Member, American Society of Mechanical Engineers.

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ROBERT W. KEATEN Business Address: GPU Service Corporation 100 Interpace Parkway Parsippany, New Jersey 07054 Education: B.S., Physics, Yale University, 1957.

Post-Graduate and Professional Courses in Mathematics, Engineering and Business, UCLA, 1960-1972.-

Ex per ience: Manager, Systems Engineering Depart-ment, GPU Service Corporation, April 1978 to present. Responsible for the development and application of specialized analytical skills in such areas as nuclear core reloads and fuel management; plant dynamic and safety analysis; system generating plant process computers; control and safety systems analysis, and analysis of plant operating performance for nuclear and fossil plants. Served as Deputy Director of Technical Support at Three Mile Island during the post-accident period.

Program Manager, Light Metal Fast Breeder Reactor Technology, Atomics International Division of Rockwell International, 1974 to 1978. Managed research and development programs performed for U.S. Department of Energy, including programs in reactor physics, safety and component development.

Manager of Systems Engineering, Light Metal Fast Breeder Reactor Program, Atomics International Division of Rockwell International, 1968 to 1974.

Responsible for performance of safety analyses, development of safety criteria and development of instru-mentation, control and safety systems decign.

American Representative to the CECD Halden Reactor Project in Norway, 1965-1968. Participated in research on nuclear fuel performance, appli-cation of digital computers to nuclear reactors, and on development and application of in-core instru-mentation.

Supervisor of Engineering, Sodium Reactor Experiment, Atomics International, Division of Rockwell International, 1962-1965.

Responsibilities included analysis and measurement of the nuclear heat transfer and hydraulic parameters of the reactor core and process systems; specification and installation of nuclear and process instrumentation; design and installation of new control systems.

Senior Physicist, Sodium Reactor Experiment, 4tomics International, Division of Rockwell International, 1959-1962. Performed measurements and analyses of the nuclear and thermal parameters of the reactor.

Experimental Physics Group, DuPont Savannah River Plant, 1957-1959.

Performed measurements and calcula-tions of the nuclear parameters of the reactor lattices.

Honors and Professional Affiliations: Member of the Nuclear Power Plant Standards Steering Committee of the American Nuclear Society.

Member and past Chairman of the LMFBR Design Criteria (ANS-54) Standards Committee of the American Nuclear Society.

Registered Professional Engineer (Nuclear Engineering) , California.

Publications: " Analysis of TMI-2 Sequence of Events Operator Response ," presented to a special session of the American Nuclear Society Conference, San Francisco, November 1979; and to Edison Electric Institute Conference, Cleveland , October 1979.

"The Role of Instrumentation in the TMI-2 Accident," presented at the American Nuclear Society Conference, June 1980.

Safety and Environmental Aspects of Liquid Metal Fast Breeder Reactors" 35th Annual American Power Conference, Chicago , Ill . , May 1973.

" Safety Aspects of the Design of Heat Transfer Systems in LMFBR's" International Conference on Engineering of Fast Reactors for Safe and Reliable Operation, Karlsruhe, Germany, October 1972.

" Safety Criteria and Design for an FBR 3emonstration Plant," ASME Nuclear Engineering Conference at Palo Alto, Calif., March 1971.

" Evaluation of Thermocouples for Detecting Fuel Assembly Blockage in LMFBR's," American Nuclear Society Annual Meeting, Los Angeles, California, June 1970.

"A Mathematical Model Describing the Static and Dynamic Instability of the SRE Core II," Reactor Kinetics and Control, AEC Symposium Series 2.

( Also published as NAA-SR-8431. )

" Reactivity Calculations and Measurements at the SRE," ANS Topical Meeting: Nuclear Performance of Power-Reactor Cores, September 1963.

" Measurement of Dynamic Temperature Coefficients by Forced Oscillations in Coolant Flow," Trans-American Nuclear Society 5, No. 1, June 1962.

" Analysis of Power Ramp Measurements with an Analog Computer," Trans-American Nuclear Society 5, No. 1, June 1962.

" Reflected Reactor Kinetics,"

NAA-SR-7263.

Many other reports covering analytical and experimental work.

MICHAEL J. ROSS Business Address: Metropolitan Edison Company Three Mile Island Nuclear Station -

P.O. Box 480 Middletown, Pennsylvania 17057 Education: U.S. Navy Nuclear Power School, 1: . U.S.

Navy Nuclear Power Prototype Schoo; 1961.

Experience: Supervisor of Operations, Three Mile Island Unit 1, Metropolitan Edison Company, 1978 to present. Responsible for directing the day-to-day operation of the plant to ensure compliance with the conditions of the plant operating license and technical spe-cifications, including supervision of the Radioactive Waste Processing and Shipment Group and coordination of operations and related maintenance activities with the Superintendent of Maintenance.

Shift Supervisor, Three Mile Island Unit 1, Metropolitan Edison Company, 1972 to 1978.

Responsible for the management of all operations and maintenance activities, including the manipulation of any controls, equipment or components in physical plant systems on his shift.

Shift Foreman, Three Mile Island Unit 1, Metropolitan Edison Company, 1970 to 1972.

Responsible for performance of various pre-operational activities, including preparation of procedures and start-up equipment checks.

Reactor Plant Technician, Saxton Nuclear Experimental Corporation,1968 to 1970.

Held position of reactor operatorr addi-

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tionally, was responsible for training operations staff.

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U.S. Navy, 1960 to 1968. Positions held include reactor operator aboard USS Haddo, Instructor at the Nuclear Power Training Unit, and AEC Field Representative at the Nuclear Power Training Unit

Professional Affiliations

Babcock & Wilcox Owner's Group, Fuel Bandling Subcommittee.

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1 LIC 11/26/80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE' ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)~

METROPOLITAN EDISON COMPANY ) Docket No. 50-289

) (Restart)

(Three Mile Island Nuclear )

Station, Unit No. 1) )

CERTIFICATE OF SERVICE I hereby certify-that copies of " Licensee's Supplemental Testimony of Robert W. Keaten, Joseph J. Colit:

and Michael J. Ross in Response to Board Question No. 6 (Emergency Feedwater Reliability)" were hand served upon those persons on the attached Service List whose names are marked by an asterisk, on the 25th day of November, and upon all others on the attached Service List by deposit in the United States mail, postage prepaid, this 26th day of November, 1980.

Thomas A. Baxter Dated:-November 26, 1980 l j

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

METROPOLITAN EDISON COMPANY ) Docket No. 50-289

) (Restart) j (Three Mile Island Nuclear )

Station, Unit No.1) )

SERVICE LIST t

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• Ivan W. Snith, Esquire John A. Invin, Esquire Chaiman Assistant 0:mnsel Atanic Safety and Licensing Pennsylvania Public Utility Ozm'n ,

Board Panel Post Office Box 3265 l U.S. Niv lane Begulatory Wi== ion Barrisburg, Pennsylvania 17120 washingtzm, D.C. 20555

• Karin W. Carter, Esquire

• Dr. Walter B. Jordan Assistant Attorney General Atanic Safety and Licensing 505 Executive House

} Pcat Offi Box 2357 Board Panel l 881 West Outer Drive h M aharg, Pennsylvania 17120 Oak Ridge, 'hnnessee 37830 John E. Minnich

• Dr. Linda W. Little Chaiman, Dauphin Cbunty Board Atanic Safet g mi Licensing of otanissicnars Board Parm Datshin County Courthouse 5000 sermitage Drive Front an$ Market Streets Raleigh, North Carolina 27612 Barrisburg, Pennsylvania 17101

• Janes R. Tourte11otte, Esquire h iter W. O) hen, Esquire Office of the Executive Imgal DiKEEx Consumr Advocate l

U. S. Nuclear Regulatory h4*=i= Office of Otmsuner Advocate m shingtcm, D.C. 20555 14th Floor, Strawberry Square Harrisburg, Pennsylvania 17127

t Docketing and Service Sectica Office of the Secretary U. S. Nuclear Begulatory Lmnissicm ,

Washington, D.C. 20555

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7 Jordan D. Cunningham, Esquire WI114mm S. Jordan, III, Esquixe Attorney for %mt Township Attorney far People Against Nuclear T.M.I. Sta-dng Ckzunittee Energy 2320 North Second Street na m & Weiss Barrisburg, Famsylvania 17110 1725 Eye Street, W.W., Suite 506 mshingtm, D.C. 20006 S eo&ne A. Adler, Esquire Midoff Ranger Selkowitz & Adlar R tert Q. Poliurd Post Office Box 1547 609 Mortp=14*' itreet Barrisburg, Pennsylvania 17105 Baltimore, Mar / land 21218

• Ellyn R. Meiss, Esquire Chauncey Fepford Attomey for the thicm of Ctocarned Judith H. Johnsru5 Scientists twiu,w.tal Coaliticm on Ev-laar Barnce & Maias Power 1725 Eye Street, N.W., Suite 506 433 Orlando Avenue mshingtm, D.C. 20006 State College, Pennsylvania 16801 Steven C. Sholly Marvin I. Imwis 304 South Market Street 6504 Bradford Terrace Mechanicsburg, Pennsylvania 17055 Philadelphia, Pennsylvania 19149 Daniel M. Pell, Esquire Marjorie M. Aamodt Anti-Nuclear Group Representing York R. D. 5 32 South Beaver Street Cbatesville, Pennsylvania 19320 York, Pennsylvania 17401 Attorney General of New Jersey Attn: Thomas J. Germine, Esq.

Gail Bradford Deputy Attorney General Anti Nuclear Group Representing Ycak 245 West Philadelphia Street Division of Law - Room 316 York, Pennsylvania 17404 1100 Raymond Boulevard Newark, new Jersey 07102 O

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