ML19260A733

Recommended Requirements for Restart of TMI-1, Vol 1. Amends 1-5 Have Been Incorporated W/Original Text

ML19260A733
Person / Time
Site:	Three Mile Island
Issue date:	09/12/1979
From:	METROPOLITAN EDISON CO.
To:
References
NUDOCS 7912030174
Download: ML19260A733 (300)

Text

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REPORT IN RESPONSE TO NRC STAFF 1

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TI'REE MILE ISIAND NUCLEAR STATIQ4 UNIT 1 RESTARP REPORP

, INSTIUCTIO4S

'Ihe attached material ccupletely replaces all previous 'IMI-l Festart Peport text. 7cundnents 1 through 5 have been incorporated witn the original text.

Please rtriove and discard all material fran the Peport In Pasponse to NPC Staff Pecomiended Pequirements for Restart of 'Ihree Mile Island Nuclear Station Unit 1 binder and replace it with the attached text. A'.so renove and replace the cover and spira with the enclosed dark red ones. ?dditional binders with further instructi:ns will be forwarded in the near fra:re.

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Po;t Office Box 480 Middletown, Pennsylvania 17o57 717 944-4041 File: 2259.10 September 12, 1979 E&L-1778 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Conmission Washington, DC 20555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION UllIT 1 (TMI-1)

DOCKET NUMBER 50-28.9 OPERATING LICENSE NO. DPR-50 Enclosed are sixty (60) printed copies of a report in loose-leaf binders that address many of the recommended requirements contained in the Commission's August 9,1979 Order and Notice of Hearing for TMI-1.

Please note that eleven (11) xerox copies of exactly the same report were submitted on September 7,1979 to expedite your staff review. If you or your staff have any questions concerning this submittal, please contact Mr. Courtney W.

Smyth at (201) 263-4900, extension 236.

Very truly yours, d4%

J. G. Herbein f

Vice President-Nuclear Operations JGH:bar Enclosure cc: Mr. J. Collins Mr. L. Engle (w/o enclosure)

Mr. R. Reid I456 248 MetrcooMan Ecson Company is a F/emoe o+ rne Genera' Puon: Ut l' ties System

Metropolitan Edison Company ay [I" ) Post 0+$ce Box 542 Readmg Pennsylvania 19640 215 92E3601 File: 2259.10 Wester's oerect osa' Nur ce' October 9, 1979 ESL-1840 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION UNIT 1 (TMI-1)

DOCKET NUMBER 50-289 OPERATING LICENSE NO. DPR-50 AMENDMENT NO. 1 TO THE 'DfI-1 RESTART REPORT Enclosed are sixty (60) copies of the subject amendment .'or incorporation into the 60 copies of the Restart Report that were submitted to you on September 12, 1979. This amendment provides additional information and revises socc information previously submitted.

We anticipate filing another amendment which will provide significant additional information related to the recommendations of NUREG-0578 as supplemented by you (RCS high point venting) and the ACRS. This amendment will be forwarded the week of October 22, 1979. In addition to the above amendment, we expect to file an amendment by the end of October which provides substantial information related to Emergency Planning.

Very truly yours,

/b

,/hvJ.G.Herbein

/ Vice President - Nuclear Operations JGH/JRT/ab Enclosures cc: J. Collins R. W. Reid H. Silver R. H. Vollmer Ue recoca- Ec; son Company is a Memee' o' t"e Gene a No": Ur met System

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2259.1 a# r.u-o.0 E&L-1842 October 10, 1979 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

THPIE MILE ISLAND NUCLEAR STATION UNIT 1 (TMI-1)

DOCKET NUMBER 50-289 OPERATING LICENSE NO. DRP-50 AMENDMENT NO. 2 TO THE TMI-l FISTART REPORT Enclosed are sixty (60) copies of the subject amendment for incorporation into the 60 copies of the Restart Report that were submitted to you en September 12, 1979. This amendment provides additional in'ormation.

Very tr"ly yours, vy CL Y J. G. Hercein Vice President - Nuclear Operations Enclosures cc: J. Collins R. W. Reid H. Silver R. H. Vollmer 1456 250

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11073 gMr ) Post O*fi:e Box 542 Reading Pennsylvania 19640 215 929 3601 venter's D..ect D.at Numbe.

File: 2259.10 E4L-1894 October 26, 1979 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Co==ission Pashing:.on, DC 20555

SUBJECT:

THPIE MILE ISLAND NUCLEAR STATION UNIT 1 (TM1-1)

DOCKET NUMSER 50-289 OPERATING LICENSE NUMBER DRP-50 AMINDMENT NO. 3 TO THE TMI-1 RESTART REPORT Enclosed are sixty (60) copies of the subject amendment for incorporation into the sixty (60) copies of the Restart Report that were submitted to you on September 12, 1979. This amendment provides additional information.

Very truly,yours, D' /

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-f/t J. G. Herbein Vice President - Nuclear Operations JGHlab Enclosures cc: J. Collins R . k' . Reid H. Silver R. H. Voll=er 1456 25i O

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Metropolitan Edison Company Post Office Box 542 Reaaing Pennsylvania 19640 215 929 3601 Writers Direct Dial Nwmoer File: 2259.10 October 31, 1979 E&L-1906 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION UNIT 1 (TMI-1)

DOCKET NUM3ER 50-289 OPERATING LICENSE NUMBER DPR-50 AMENDMENT NO. 4 TO THE TMI-l RESTART REPORT Enclosed are sixty (60) copies of the subject ataendment for in-corporation into the sixty (60) copies of the Restart Report that were submitted to you on September 12, 1979. This amendment provides additional information.

Very truly yours, O

J. G. Herbein Vice President - Nuclear Operations JGH/JRT/ab Enclosures cc: J. Collins R. W. Reid H. Silver R. H. Voll=er 1456 252

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Metropolitan Edison Company

, Post O'fice Box 542 Reading Pennsylvania 19640 215 929-3601 Wrrftr's Direct Dis' NumDer File: 2259.10 Noverber 7, 1979 E&L-1922 Dr. Harold R. Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION UNIT 1 (TMI-1)

DOCKET NUMBER 50-289 OPERATING LICENSE NUMBER DPR-50 AMENDMENT NO. 5 TO THE TMI-l RESTART REPORT Enclosed are forty (40) copies of the subject amendment for incorporation into the sixty (60) copies of the Restart Report that were submitted to you on September 12, 1979. An additional twenty (20) copies will be forwarded separately. This amendment provides additional information in response to your letter of October 26, 1979.

Very trulf yours,

/b M J. G. Herbein Vice President -

Nuclear Operations JGH/ab Enclosures cc: J. Collins R. W. Reid H. Silver R. H.Vollmer 1456 253 Metrecohtan Ec son Comcany rs a Memte of tne Gene a, cuoc Utnes 5fstem

TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND REPORT ORGANIZATION 1-1 1.1 Introduction 1-1 1.2 Report Organization 1-1 1.3 Abbreviations 1-2 2.0 PLANT MODIFICATIONS 2.1-1 2.1 General 2.1-1 2.1.1 Short-Term Modifications 2.1-1 2.1.1.1 Reactor Trip or Loss of Feedwater/ 2.1-3 Turbine Trip 2.1.1.2 cvsition Indication for PORV and Safety 2.1-4 Valv e s 2.1.1.3 Emergency Power Supply Requirements 2.1-5 for Pressurizer Heaters, PORV, Block Valve , and Pressurizer Level Indication 2.1.1.4 Post LOCA Hydrogen Recombiner System 2.1-8 2.1.1.5 Containment Isolation Modifications 2.1-11 2.1.1.6 Instrumentation to Detect Inadequate 2.1-17 Core Cooling 2.1.1.7 Auxiliary Feedwater Modifications 2.1-20 2.1.2 Long-Term Modifications 2.1-27 2.1.3 Met-Ed Initiated Modifications 2.1-28 3.0 PROCEDURAL MODIFICATIONS 3-1 3.1 General 3-1 3.1.1 Emergency Procedures 3-2 3.1.2 Administrative Procedures 3-2 3.1.3 Surveillance / Preventative Maintenance / Corrective 3-3 Maintenance Procedure 3.1.4 Operating Procedures 3-3 i Am. 3 1456 254

_ TABLE OF CONTENTS _ Continued

_P, a_ge 4.0 EMERGENCY PLANNING 4-1 4.1 Introduction 4-1 5.0 THREE MILE ISLAND NUCLEAR STATION ORGANIZATION 5-1 5.1 General 5-1 5.2 Station organization 5-2 5.2.1 Vice President - Nuclear Operations 5-2 5.2.2 Unit Superintendent 5-2 5.2.3 Supervisor of operations 5-2 5.2.4 Shift Supe rv iso r 5-2 5.2.5 Shif t Foreman 5-3 5.2.6 Supervisor Preventative Maintenance 5-3 5.2.7 Director - Technical Support 5-3 5.2.8 Shift Technical Engineer 5-3 5.2.9 Manager of Support Services and Logistics 5-3 5.2.10 Supervisor - Radiation 5-4 5.2.11 Manager - Training 5-4 5.3 Station Support Organization 5-4 5.4 Minimum Qualification Requirements for IMI Unit 1 Personnel 5-6 5.4.1 Vice President - Nuclear Operations 5-6 5.4.2 Unit Superintendent 5-6 5.4.3 Supervisor of Operations 5-7 5.4.4 Shif t Supervisor 5-7 5.4.5 Shift Foreman 5-8 5.4.6 Supervisor Pre /entativ e Maintenance 5-9

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TABLE OF CONTENTS - Continued Fage 5.4.7 Director - Technical Support 5-10 5.4.8 Shift Technical Engineer 5-10 5.4.9 Manager of Support Services and Logistics 5-12 5.4.10 Supervisor - Radiation Protection and Chemistry 5-12 5.4.11 Manager - Training 5-12 6.0 OPERATOR ACCELERATED RETRAINING PP0 GRAM 6-1 6.1 Introduction 6-1 6.2 Program Objectives 6-1 6.3 Topical outline 6-2 6.4 Program Rationale 6-5 6.5 Instructional Procedure 6-6 6.6 Evaluation Procedure 6-9 6.7 Program Format 6-11 7.0 RADWASTE MANAGEMENT 7-1 7.1 General 7-1 7.1.1 Near Term Modifications 7-1 7.1.2 Long Term Modifications 7-2 7.2 Discussion of Specific Items 7-2 7.2.1 Specific Areas of Separation / Isolation 7-2 7.3 Radwaste Capability 7-6 8.0 SAFETY ANALYSIS 8-1 8.1 Introduction 8-1 8.2 Areas of Investigation 8-2 iii 1456 256 im. 2

TABLE OF CONTENTS - Continued Page 8.2.1 Modifications Resulting from the 8-2 August 9, 1979 Order 8.2.2 Modif'. cation as Result of Order of May, 8-2 1978 8.2.3 Modification Originating from within Met-Ed 8-3 8.2.4 I&E Bulletin 79-05C 8-3 8.3 Effect of Changes on Safety Analysis 8-3 8.3.1 Rod Withdrawal from Startup 8-3 8.3.2 Rod Withdrawal at Power 8-8.3.3 Moderator Dilution Accident 8-5 8.3.4 Cold Water Addition 8-5 8.3.5 Loss of Coolant Flow 8-6 8.3.6 Dropped Control Rod 8-7 8.3.7 Loss of Electric Power 8-8 8.3.8 Station Blackout (Loss of AC) 8-8 8.3.9 Steam Line Failure 8-9 8.3.10 Steam Generator Tube Failure 8-11 8.3.11 Fuel Handling Accident 8-12 8.3.12 Rod Ejection Accident 8-12 8.3.13 Feedwater Line Break Accident 8-13 8.3.14 Waste Gas Decay Tank Rupture 8-14 8.3.13 Small Break Loss of Coolant Accidents (LOCA) 8-15 8.3.16 Large Break Loss of Coolant Accidents (LOCS) 8-18 8.4 Summary and Conclusions 8-19 9.0 DRAWINGS 9-1 1456 257 iv Am. 3

TABLE OF CONTENTS - Continued Page 10.0 CROSS REFERENCE TO ORDER RECOMMENDATIONS 10-1 10.1 In troduc tion 10-1 10.2 Short-Term Recommendations and Met-Ed Responses 10-1 10.3 Specific Responses to Recommendations 10-4 10.3.1 Response to IEB 79-05A, Item 2 10-4 10.4 Long-Term NRR Recommendations and Met-Ed Responses 10-6 11.0 _ TECHNICAL SPECIFICATIONS 11-1 11.1 Introd uc tion 11-1 11.2 Technical Specification Changes 11-1 11.2.1 Auxiliary (Emergency) Feedwater ( AFW) 11-1 11.2.2 Reactor Trip on Loss of Feedwater or Turbine 11-1 Trip 11.2.3 High Pressure Trip Setpoint Reduction 11-1 11.2.4 Containment Isol.tior Se tpoint s 11-1 11.2.5 Hydrogen Recombiner 11-2 11.2.6 TMI-1/IMI-2 Separation 11-2 11.2.7 Administrative Controls 11-2 1456 258

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1.0 INTRODUCTION

AND REPORT ORGANIZATION

1.1 INTRODUCTION

Metropolitan Edison Company (Met-Ed) applied for a license to construct and operate Three Mile Island Nuclear Station Unit 1 (TMI-1) on May 1, 1967 (TMI-l is jointly owned by Met-Ed, Jersey Central Power and Light (JCP&L), and Pennsylvania Electric Company (Penelec) but operated by Met-Ed. Me t-Ed , JCP6L and Penelec are wholly owned subsidiaries of General Public Utilities (GPU).) Following issue of the Atomic Energy Commissions ( AEC)

Safety Evaluation Report (February 5,1968 as supplemented April 26, 1968) and hearings before the Atomic Safety and Licensing Board ( ASLB) the AEC issued a permit to construct TMI-l (CPPR-40) on May 18, 1968.

On March 2, 1970, Met-Ed filed the Final Safety Analysis Report (FSAR) and Operating License Application for TMI-1. The applica-tion was for operation at a core power level of 2535 megawatts thermal (MWt) based on Babcock & Wilcox (B&W) analyses performed for a core power level of 2568 MWt. Based on its SER issued June 11, 1973, the AEC issued Operating License DPR-50 on April 19, 1974.

TMI-l achieved initial criticality on June 5,1974 and was declared " Commercial" on September 2, 1975. Since commerciel operation TMI-1 has been refuelled five times. The unit was ready to begin operation on the fifth core on March 28, 1979 when the TMI-2 accident occurred. Until cenditions at TMI-2 were fully understood Met-Ed decided to keep TM!-l shutdown. On April 16, 1979, Met-Ed committed to providing the NRC with significant advance notice prior to startup of TMI-1.

On June 28, 1979, Met-Ed informed the NRC that TMI-l would not be started up until certain plant modifications were completed. The NRC issued an Order on July 2,1979 that TMI-1 remain sht tdown until after a public hearing and further commission order. The Commission issued a further Order and Notice of Hearing on August 9,1979 which included a list of requirements which the Director of NRR had recommended as a condition for restart of TMI. This report addresses these recommended requirements, except that the requirements for a demonstration of managerial capability and financial resources and of financial qualifications will be separately addrersed.

1.2 REPORT ORGANIZATION This report is composed of eleven (11) sections which, combined, cover the August 9,1979 Order requirements. All requirements of a related nature are discussed in a single section. For example all requirements related to plant hardware modifications are presented in Section 2 and referenced by other Sections as appropriate. Section 10 provides a discussion of how a require-ment is met or where in the report the discussion can be found.

1-1

1.3 ABBREVIATIONS Abbreviations or Acronyms are frequently used throughout this report. The ones more commonly used are defined below:

ACRS Advisory Committee on Reactor Safeguards B&W Babcock & Wilcox CRDM Control Rod Drive Mechanism DH Decay Heat ECCS Emergency Core Cooling System ES Engineered Safeguards FSAR Final Safety Analysis Report HPI Righ Pressure Injection ICS Integrated Control System LOCA Loss of Coolant Accident LPI Low Pressure Injection MU Makeup NPSH Ne t Positive Suction Head NRC Nuclear Regulatory Commission PORV Power Operated Relief Valve PRZR (PZR) Pressurizer psig pounds per equare inch gauge QA Quality Assurance RB Reactor Building RCDT Reactor Coolant Drain Tank RCP Reactor Coolant Pump RCS Reactor Coolant System SFAS (ESFAS) Safety Features Actuation System TMI Three Mile Island 1.4 DEFINITIONS Safety Grade - Safety grade within the context of this report means that a system or cumponent has (unless otherwise stated) the following features: redundancy, testaoility, reliable onsite power source, and capability to withstand appropriate adverse environments (including seismic).

Safety Related (Nuclear) - Safety related means that the system or component is useful in protecting nuclear safety. Nuclear safety related items may be safety grade or non-safety grade. In general only those items which are nuclear safety related and form the primary line of defense (in a defense in epth approach to safety) are safety grade.

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2.0 PLANT MODIFICATIONS 2.1 GENERAL The plant design features related to safe operation have been described in detail in the TM;-l FSAR and in various submittals to the NRC since the issuance of TMI-l's Operating License on April 19, 1974. Further modifications are being made to the plant in response to the staff's recommendations contained in the Commission's Order dated August 9,1979. These modifications are described below and will be completed before startup of IMI-l or shortly thereafter. Modifications to be completed before startup are described Section 2.1.1 and modifications which may be completed later are described in 2.1.2. In addition Section 2.1.3 describes certain additional modifications not included in the staff's recommendations. The,e modifications were proposed by Met-Ed in its June 28, 1979 letter to NRC and are to be completed prior to restart of Til-1.

2.1.1 Short-Term Modifications 2.1.1.1 Reactor Trip or Loss of Feedwater/ Turbine Trip 2.1.1.1.1 System Description A Safety Grade Reactor Trip System will be installed before

• restart in order to implement a reactor trip upon loss of both main feedwater pumps or upon turbine trip. This system will derive four channel signals for loss of feedwater pumps and turbine trip. They will be used as inputs to the existing Reactor Protection System and will result in a trip of the reactor on coincident two of four signals.

2.1.1.1.2 Design Bases The Safety Grade Reactor Trip System is designed to provide a reactor trip upon loss of main feedwater pumps or a turbine trip as an anticipatory trip. This would preclude reactor trips on high pressure for the anticipated transient conditions and minimize the challenges to the Pressurizer PORV and safety valves. The system will use redundant, four channel signals corresponding to those of the existing Reactor Protection System.

It will be testable and meet the single failure criterion of IEEE-279.

2.1.1.1.3 System Design A diagram of the system is shown in Figure 2.1-1. The turbine trip and feedwater pump trip signals will be derived f rom pres-sure switches on the control oil systems of the main turbine and the FW pump turbines respectively. Each of the four channel signals will be connected to the corresponding channel of the Reactor Protection System. The signals will be fed through 2.1-1 Am. 3 1456 263

.g contact isolators to preclude the propogation of faults into the W Reactor Protection System (RPS). The reactor will be tripped through the existing RPS logic upon coincident signals from any two of the four channels.

A bypass arrangement will be provided in order to allow for power escalation, starting the main turbine and normal shutdown of the main turbine. The main turbine trip bypass will be automatically placed in effect when reactor powe: is less than 20%. The bypass will be automatically removed when the reactor power is increased above 20%. Bypass of the feedwater pump trip signal is automa-tically placed in effect when reactor power is less than 10%. It will be removed automatically when reactor power is raised above 10%. The bypass fnetion will be accomplished individually in each of the four channels by means of bistables which m)nitor the power range nuclear instrumentation.

The additional modules required in the Reactor Protection System will be the same safety grade equipment type used in the original system. Wiring for redundant channels will be separated and run in Seismic I, safety grade raceways except in the turbine build-ing. Since the turbine building is not Seismic I, the equipment and wiring therein cannot be classified as Seismic I. Howev e r ,

all wiring in the turbine building for this system will be run in conduit and redundant channgls will be routed separately to minimize the probability of disabling more than one channel due to damage to the turbine building. The system will be designed with normally closed contacts so that an open wire will represent a tripped condition. The signals from the turbine building will go through contact isolators in the Reactor Protection System to preclude the propogation of faults in the system.

2.1.1.1.4 Design Evaluation The Safety Grade Reactor Trip scheme provides an anticipatory trip to tte reactor, reducing the number of reactor trips on high pressure ind the number of challenges to the Pressurizer PORV and safe ty vrives.

2.1.1.1.5 Safe ty Evaluation The system is safety grade and meets the requirements of IEEE-279 including those for testability and single failure criterion.

The modifications which will be required to the Reactor Protec-tion System will not degrade the ability of that system to perform its design function. The design will result in an enhancement of nuclear safety.

2.1.1.1.6 Start-Up Testing This system will be tested during installation to verify its operation prior to start-up.

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2.1.1.2 Position Indication for PORV and Safety Valves 2.1.1.2.1 System Description The purpose of this modification is to provide the Control Room Operator with information on the status of the pressurizer electromatic relief valve RC-RV2 and the pressurizer code safety valves RC-RVI A and RC-RVlB. Discharge flow will be measured by differential pressure transmitters connected across elbow taps downstream of each of the valves. In addition, the electromatic relief valve will be monitored by accelerometers mounted on the valve. These will detect flow if the valve opens. Alarms and indications will be provided in the control room to inform the operator if any of these valves are open.

2.1.1.2.2 Design Bases A reliable and unambiguous control grade indication wi).1 be provided to the Control Room Operator if the pressurizer elec-tromatic relief valve or code safety valves open. The monitoring system will remain functional in containment conditisas associat-ed with any transient for which valve status is required by the operator. Redundant and diverse means will be provided for monitoring the electromatic relief valve (RC-RV2). The monitor-ing systems will remain functional during a loss of off-site powe r. All equipment inside containment will be seismically mounted. The integrity of existing safety related systems will not be impaired by this modification.

2.1.1.2.3 System Design All of the system components have been selected for reliable operation and, where applicable for operation under adverse conditions inside containment. The differential pressure trans-mitter model which has been selected has previously been quali-fled for operation in a post-LOCA environment, fa- ability to operate af ter a seismic event and withstand 2.2 > (08 rads.

The components comprising the acoustic flow detect or have been previously used by B&W in the Loose Parts Monitoring System.

They have been seismically tested and have been tested under the B&W " Steam Line Break" and "Small Break LOCA" containment environ-ments. They will withstand 108 rads. The monitoring systems will be supplied from on-site electrical power supplies. Dive rse and redundant means will be used for monitoring of the electro-matic relief valve. Both differential flow measurement and acoustic detectors will be provided.

2.1.1.2.4 Design Evaluation Elbow taps are widely used for flow measurement in fluid systems and a great deal of empirical data is available for calculating expected dif ferential pressure across elbow taps for given flow conditions. Calculations have been made, using conservative assumptions, to demonstrate that a satisfactory signal wil . be 2.1-3 1456 265 3

generated when any of the valves open. Calculations have been made for saturated, liquid and two phase flow. A summary of these calculations is provided in Appendix 2A. Tests run by B&W on the electromatic relief valve under reduced flow condiitons have confirmed the validity of this approach. Because of the straight-forward and well known relationships that exist between flow conditions and differential pressure across the elbow, the signal from one differential pressure transmitter can be confi-dently predicted for any flow conditions. For this reason it has been concluded that operating tests, which would be difficult since they involve opening the PORV and relief valves, will not be required.

Acoustic monitoring of the electromatic relief valve makes use of well proven equipment and techniques which have been used in the B&W Loose Parts Monitoring System. Tests run on this valve at the B&W Alliance facility demonstrated that the acoustic monitor-ing system gave satisfactory results.

2.1.1.2.5 Safety Evaluation Instrument taps will be installed on elbows in the discharge piping of pressurizer code safety valves RC-RVIA and RC-RVlB and electromatic relief valve RC-RV2. This piing is classified as N2, Seismic I. Analysis has been performed to demonstrate that this modification will not degrade the integrity of the existing pipe . The pipe classification has been maintained up to and including the instrument root valves. The mounting of new equipment which will be located in the vicinity of safety related systems has been analyzed to ensure that no hazardous missiles will be generated in a seismic event. It has been concluded that this modification will not degrade any safety related systems.

The valve position indication f unction has not been classified as safety grade.

2.1.1.2.6 Instrumentation The output signals from the three differential pressure trans-mitters will be displayed on indicators in the control room.

They will be calibrated in " inches of water". Each signal will also go to an alarm bistable. A control room alarm will be initiated if any of the signals exceed a pre-determined value.

This will alert the operator that one of the valves is open. The differential pressure signal will also be monitored by the plant computer for logging, trending, and alarm functions.

The outputs from the accelerometers which will be mounted on RC-RV2 will be processed by monitoring equipment installed in the existing Loose Parts Monitoring Cabinet. An output signal indicative of flow through the valve will be displayed and recorded locally. A control room alarm will be initiated if flow is detected. This signal will also be monitored by the plant computer for logging, trending, and alarm purposes.

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2.1.1.3 Emergency Power Supply Requirements for Pressurizer Heaters, PORV, Block Valve, and Pressurizer Level Indication 2.1.1.3.1 Pressurizer Heaters 2.1.1.3.1.1 System Description The purpose of this modification is to provide redundant emer-gency power for the 126 KW of pressurizer heaters required to maintain natural circulation conditions in the event of a loss of offsite power. A manual transfer scheme will be installed to transfer the source o' ver for 126 KW of pressurizer heaters from the balance of plant (B0P) source to a " Red" engineered safeguards (ES) source. A similar manual transfer scheme will be installed to transfer the source of power for 126 KW of pressur-izer heaters from the BOP source to a " Green" ES source. Each manual transfer scheme will have double isolation on each end of the transfer and have mechanical key interlocks to govern the order of the transfer procedure. Figure 2.1-4 is a schematic representation of these transfer schemes.

2.1.1.3.1.2 Design Basis Babcock and Wilcox has recommended that at least 126 KW of pressurizer heaters be restored from an assured power source within two hours af ter a loss of of fsite power. Separation and isolation of Class IE equipment and circuits from non-Class IE equipment and circuits will be in accordance with Regulatory Guide 1.75.

The 480 volt ES circuit breaker is the isolating device between l

Class IE and non-Class IE portions of the design. The Class lE ,

portion of the design is that portion up to and including the 480  !

volt ES circuit breaker and protective elements. Unerv oltage l relays connected to the 480 volt ES bus will detect a fault that I is of sufficient magnitude to endanger the ability of safety l loads on the bus to start or run. The undervoltage relays will  !

initiate tripping of the 480 volt ES circuit breaker feed to the I pressurizer heaters and thereby remove any endangerment caused by that circuit.

While the remaining portion of the design is classified non-Class IE, separation will be maintained up to the pressurizer heater terminal box T-161 which is located on the secondary shield wall.  ;

The remaining portion of the design (i.e. , terminal box T-161, I pressurizer heater elements and interconnecting cable routing) [

will remain as it presently exists. The constraints imposed by  !

the plant's original physical construction make this necessary.

The relative closeness of the pressurizer heater elements and heater bundles does not permit further physical separation.

1456 267 2.1-5 Am. 3

The pressurizer heaters are passive devices. The only common ev ent to affect both redundant emergency pressurizer heater circuits could only occur in that area between terminal box T-161 and the pressurizer heater elements. For such an event, the protection on the 480 volt ES feed to the pressurizer heaters will be fully coordinated with the protection on the 480 voit ES bus main circuit breaker.

If both emergency circuits to the pressurizer heater are lost, reactor coolant pressure shall be sufficiently maintained through use of the make-up (HPI) pumps.

The double isolation in the form of the circuit brekers and removable elements (see Figure 2.1-4) along with the kirk key interlocks, preclude lining up the system with a 480 volt ES bus connected to the BOP bus or lining up the system with the " Red" 480 volt ES bus connected to the " Green" 480 volt ES bus.

B&W determined the number of pressurizer heaters by taking into account the following:

1. The loss through the pressurizer insulation was calculated.

The service areas of the insulation was determined and an average heat flux for the outside service area was assumed to be 80 BTU /hr ft2 This calculated to a heat loss of approximately 96,000 BTU /hr.

2. The loss through the uninsulated pressurizer areas around the horizontal heater bundles was calculated in the same manner as item 1 and resulted in an approximate heat loss of 50,000 BTU /hr.

3. B&W's experience has shown that the insulated heat losses account for less than half of the total losses. Therefore, a factor of 2.5 was applied to the sum of the accounted losses.

Thus, the total calculated heat loss from the system is 365,000 BTU /hr or 107KW. Due to the grouping at the pressurizer heaters, one bank of pressurizer heaters consisting of 126KW was recom-mended.

The time for establishing the heaters is determined by the amount of heat losses from the pressurizer and the initial water level in the pressurizer. Figure 2.1-2 shcws the expected response after establishing natural circulatiot: with no heat input from

'he heaters.

. From the Figure, two hours is sufficient time, for the heaters to operate, to insure natural circulation at hot standby after a loss of offsite power.

1456 268 2.1-Sa Am. 3

2.1.1.3.1.3 System Design Existing spare Class IE 480 volt circuit breakers on the " Red" and " Green" ES systems will be utilized for the two transfer schemes. The removable element assemblies for each transfer scheme will consir.t of two cabinets and one tab-keyed, removable element. One sabinet will be located near and connected in series with the 480 V ES circuit breaker. The other cabinet will be located near cnd connected in series with the Pressurizer Heater Control Center circuit breaker. Class lE qualified power cable will connec the load sides of the disconnect switches as shown in Figure 2.1-4. Class IE qualified under-voltage relays will be installed on each ES bus. They will initiate tripping of the ES circuit breaker to the pressurizer heaters when the bus voltage drops belaw its set point. The set point will be chosen so that starters on the ES bus can pickup if energized and the voltage at the ES motors is not lower than their ratings allow.

An Engineered Safeguards actuation signal shall trip but not lockout each ES circuit breaker to the pressurizer hecters. The remainder of the electrical power distribution system to the pressurizer heaters will remain as it presently exists.

2.1.1.3.1.4 System Operation All pressurizer heaters will be powered f rom the BOP electrical power distribution system when of fsite power is available. Upon a loss of offsite power, manual transfers will enable each of the onsite emergency diesel generators (" Red" and " Green") to provide power to 126 KW of pressurizer heaters when the diesel generators can accommodate that load. Procedures will call for tripping non-essential loads to accomplish this within the two-hour requirement. Mechanical key interlocks will dictate that the order of events in the transfer from BOP to ES power source will be as follows:

A. Opening the circuit breaker in the PHCC which will allow removal of key #1.

B. Key #1 will open the cabinet door of the disconnect switch located near the PHCC. The removable element will then be removed along with Key #2 and carried to ti e 463 V ES switch-room.

C. The removable element will be inserted into the appropriate cabinet. Key #2 will lock that cabinet door and allow removal of Key #3.

D. Key #3 will remove the inhibit feature from the 480V circuit breaker.

1456 269 2.1-6 Am. 3

E. The circuit breaker control switch will then be operated to close the ES circuit breaker feed to the transferred pres-surizer heaters when it has been established that bus loading and emergency D/G loading permit doing so.

When offsite power is restored, the reverse procedure will be used to transfer back to the BOP source.

2.1.1.3.1.5 Safety Evaluation The manual transfer scheme design provides double Class IE separation of the ES system from the BOP system - the ES circuit breaker and the removable element. Taking into account the single failure criteria, f aults on the B0P system will, at most, cause the loss of one 480 volt ES system. The transfer scheme design also precludes the connection of the " Green" ES system to the " Red" ES system.

2.1.1.3.1.6 Inservice Testing Requirements The emergency diesel generator loading procedure will be rewrit-ten to incorporate this modification. Therefore, these transfer schemes will be tested when the emergency diesel generators are tested.

2.1.1.3.2 Power Operated Relief Valve (PORV)

The present plant design is such that emergency diesel generator power will be supplied to the PORV (RC-V2) upon loss of of fsite powe r. The PORV is powered from the 250 VDC Distribution Panel IC which in turn is pwoered from the " Red" and " Yellow" ES batteries and ES Battery Chargers IA, 1C, and IE.

2.1.1.3.3 Block Valve The present plant design is such that emergent.y diesel generator power will be supplied to the block valve (RC-V3) upon loss of offsite power. The block valve is powered fr om the 480 V En-gineered Safeguard Valve Conrol Center IC.

2.1.1.3.4 Pressurizer Level Instrumentation The present plant design is such that emergency diesel generator power will be supplied to the pressurizer level instrumentation power supplies (RC-1-LT1, RC-1-LT2, RC-1-LT3) upon loss of offsite power. The pressurizer level instrumentation power supplies are parc of the ICS, NNI System, and are powered from the 120 volt ICS, NNI Power Distribution Panel ATA. That panel is, in turn, powered from the 120 volt Vital Distribution Panel VBA.

1456 270 2.1-7 Am. 3

2.1.1.j.* Pressurizer Lev el Instrumentation The present plant design is such that emergency diesel generator power will be supplied to the pressurizer level instrumentation power supplies (RC-1-LT1, RC-1-LT2, RC-1-LT3) upon loss of of f site power. The pressurizer level instrumentation power supplies are part of the ICS , NNI System, and are powered from the 120 volt ICS, NNI Power Distribution Panel ATA. That panel is , in turn, powered frow. the 120 volt Vital Distribution Panel VB A.

1456 271 2.1-7a Am. 1

2.1.1.4 POST LOCA HYDROGEN RECOMBINER SYSTEM 2.1.1.4.1 System Description The purpose of this modification is to provide a system which shall serve as a means of controlling combustible gas concentra-tions in containment following a loss of coolant accident (LOCA).

After a LOCA, the containment atmosphere of a PWR is a homo-geneous mixture of steam, air, solid and gaseous fission products, hydrogen and water droplets containing boron, sodium-hydroxide and/or sodium thiosulfate. During and following a LOCA, the hydrogen concentration in the containment results from radiolytic decomposition of water, zirconium-water reaction and aluminum reacting with the spray solution.

If excessive hydrogen is generated it msy combine with oxygen in the containment atmosphere. The capability to mix the combus-tible atmosphere and prevent high concentrations of combustible gases in local areas is provided by the reactor building ventila-tion system. The hydrogen combiner system must be capable of redecing the combustible gas concentrations within the contain-ment to below 4.1 volume percent.

The recombiner shall be capable of removing containment air mixed with hydrogen, recombine the hydrogen and exhaust the processed

. air back into the containment. This system is not required during normal plant operation.

2.1.1.4.2 Design Basis The recombiner system shall meet the design and quality assurance requirements for an engineered safety feature in terms of redun-dancy for active components, electrical power and instrumentation.

The design bass for the system shall be a loss-of-coolant accident (LOCA) with hydrogen generation rates calculated in accordance with NRC Regulatory Guide No. 1.7.

The hydrogen recombiner to be utilized for the system shall be the Rockwell International, Atomics International Div. recombiner unit purchased for TMI Unit No. 2.

One hydrogen recombiner will be installed prior to restart. The second (redundant) recombiner need not be installed, however, the piping system, electrical power supplies and structural provisions shall be installed and available. The second hydrogen recombiners shall be installed after an accident within the time period available before they need to be operational.

2.1.1.4.2 Tne system will be designed to meet the criteria of NRC Regula-tory Guide 1.7, the acceptance criteria of SRP 6.2.5, NUREG 0578 (July 1979), 10CFR50 Appendix A-General Design Criteria for containment design and integrity and 10CFR100 Reactor Site Criteria for limits of offsite releases.

2.1-S

2.1.1.4.3 System Design The system design provides an installed and a location with installed piping for a future redundant hydrogen recombiner. The recombiners will be located in the Intermediate Building at floor elevation 305 f t. , in the Leak Rate Test equipment area, as shown in Fig. 2.1-6. This system will utilize the existing "Contain-ment Vessel Leak Rate test" penetrations (nos. 415 and 416) as shown diagramatically in Fig. 2.1-7.

Since only active component failure needs to be considered, common containment penetrations will be utilized for the redun-dant recombiners. All active components will be redundant and will be provided with independent power supplies.

All system components forming the containment boundary will meet the containment isolation criteria and will be designed to Jafety Class 2 per ANSI B-31.7. All system supports will be design-ed for the DBE as seismic class S-I. The recombiners will be powered from Class IE power sources. The inside containment isolation valves will be solenoid, de power, operated valves.

The recombiner cooling air will be discharged directly to the outside environment. An evaluation will be performed to demon-strate that potential releases of intermediate building air used for recombiner cooling will not result in off site releases in excess of 10CFR100.

2.1.1.4.4 System Operation The system is designed to maintain the hydrogen concentration inside containment below the 4.1 percent by volume , lower flam-mability limit of hydrogen.

Based on the hydrogen generation rate calculated in accordance with NRC Reg. Guide 1.7, the hydrogen recombiner should start processing the containment gases when the hydrogen concentration reaches 3 percent by volume of the total containment.

The recombiner is placed into operation by opening the contain-ment isolation valves af ter having sampled the containment atmosphere and then turning on the recombiner from its remote-local panel. Local monitoring of the control panel is required until the reaction chamber reach <.s the required temperature for a self sustaining reaction between hydrogen and oxygen. Once the system is in a recombination mode, only periodic inspection at the control panel is required. A single remote recombiner alarm is provided in tne main control room to advise the operator of an operating problem with the recombiner.

When the hydrogen concentration has dropped to an acceptable level, the system is shutdown and the containment isolation valves are closed.

2.1-9

2.1.1.4.5 Safety Evaluation The hydrogen recombiner system is designed as a nuclear safety class 2, sesimic class S-I system with class IE power supply.

Containment integrity is normally maintained by double valve isolation (with a valve inside and another outside containment).

While the recombiner is being utilized for post-LOCA hydrogen control, containment integrity a; the penetration is maintained by a single, manually operated, locked closed valve located outside of containment and the redundant isolation is provided by a blind flange also located outside containment.

In order to insure the ability to draw and return containment atmosphere, considering single active failure of the power operated inside containment isolation valve , two such valves are provide per penetration with each of a redundant pair of valves powered fromalternate de pwoer supplies. These isolation valves are designed to fail closed on loss of power in order to maintain containment integrity.

All other active components have redundancy by virtue of the redundant recombiner skid and control panel. Each panel may be powered by either the " Red" or " Green" Engineered Safeguards System power supply.

Of f site releases due to leakage and discharge to the atmosphere with the recombiner cooling air will be evaluated to demonst: Ate these releases to be below the 10CFR100 limits.

2.1.1.4.6 Inservice Testing Requirements No inservice testing is required for the Hydrogen Recombiner System. However, normal inspection, testing and maintenance will be performed in accordance with standard plant operating proced-ures.

1456 274 2.1-10

2.1.1.5 Containment Isolation Modifications 2.1.1.5.1 System Description The functional requirements of. the additional containment isola-tion signals are the following:

1. Provide diverse containment isolation signal from the appli-cable reactor trip, high radiation, 1500 psig SFAS, or pipe break signal. These signals will assure that radioactive material is not transferred out of the reactor building before a 4 psig isolation signal is reached.

2. All lines open to the contrinment atmosphere or connected directly to the RCS (either normally or intermittently which can result in transfer of radioactivity outside containment),

which are neither part of the Emerg'ency Core Cooling Systems nor support for RCP operation, should be isolated on reactor trip.

3. In order to maintain non-ECCS support services for RCP opera-tion, t..e following service lines should be classified as Seismic Category I and closed on the foll wing signals, provided that the piping is protected from pipe whip and/or jet impingement (see Fig. 2.3-5), Deletion of 4 psig RB Isolation Signal Logic):

a. Reactor coolar.t pump seal return should be isolated on 30 psig reactor building pressure signal or by the operator

hrough remote manual operation on high radiation alarm.

b. Nuclear Services Closed Cooling (NSCC) water and Interme-diate Closed Cooling (ICC) water should be isolated in accordance with the logic of Figure 2.1. The only exception is that the ICC supply to the CRDM coolers shall be isolated by the 4 psig reactor building pressure signal.

c. Normal fan cooler coils will be isolated by 4 psig reactor building pressure signal and 1500 psig SFAS.

Emergency cooling will be initiated by the 1500 psig signal.

In order to utilize specific systems which have been auto-matically isolated , an isolation signal override capability is required. The isolation signal override shal) be either on a total basis or on an individual penetratiot basis dependent on the isolation signal source and the penetration which is to be opened. See Table 2.3-1 for a listing of penetrations and the required isolation override require-ments.

1456 275 2.1-11 Am. 5

The radiation monitoring shall be accomplished at the loca-tions indicated on Table 2.3-2.

High Radiation alarms shall be provided in the control room for each radiation monitor that provides a high radiation closure signal nd for the RC pump seal return line. Each alarm window shall also identify the valves which it is closing or is to be closed by the operator.

4. Specific requirements for each containment isolation valve are tabulated in attached Table 2.3-3. This table identifies the isolation signal for each valve and pipe upgrading requirements for each piping system.

5. Before the existing 4 psig reactor building pressure isola-tion signal may be deleted from the plant design, the piping system mut t be evaluated, utilizing the logic shown in attached Figure 2.3-5, to demonstrate that containment integrity will be maintained.

6. Containment isolation signal override capability will be provided in accordance with attached Table 2.3-1 which lists the following types of overrides:

a. Individual Isolation Signal Bypass - This override shall be capable of bypassing only the specific isolation signal to the appropriate valves associated with only the penetration which it is desired to open. This type of override is noted by an "I" on Table 2.3-1. The initiat-ing isolation condition may still exist when utilizing this overrida.

b. Common Isolation Signal Bypass - This override shall be a common override capable of bypassing only the specific isolation signal to all of the appropriate valves asso-ciated with the various penetrations which may be desiv J to open by the operator. The common isolation signal bypass shall also provide the override for the individual isolation signal bypass. This type of override is noted by a "c" on Table 2.3-1. The initiating isolation condition may still exist when utilizing this override,

c. Automatic Isolation Signal Bypass - The isolation signal for this type of override shall automatically be cleared although the initiating isolation condition may still exist. This will allow the operator to simply push 2.1-12 Am. 5 1456 276

the valve switches to "open" position in order to re-open the valves. This feature is used only for the RC system letdown isolation valves af ter they have been closed by a reactor trip only. This type of override is noted by an "A" on Table 2.3-1.

d. No Bypass Capability - This override shall not pe rmit the operator to re-open the valve unless the initiating condition is removed. If the isolation valves have been re-opened and the initiating condition re-eccurs then the valves shall again be isolated.

The containment isolation overrides shall be on an individual sign'l source basis such that overriding the isolation signal due to one source will still allow the valves to be isolated by a second isolation source if it is activated.

2.1.1.5.2 De sign Bases

1. The diverse containment isolation system shall meet the single failure criterion of IEEE No. 279.

2. Redundancy of sensors, measuring channels, logic, and actua-tion devices shall be maintained and not be degraded by the modifications.

3. Electrical independence and physical separation stall be in accordance with IEEE-383, where practicable. If not possible, existind physical separation criteria will be maintained.

4. Switches, independent of the automatic instrumentation, shall be provided for manual control of all containment isolation valves modified.

5. Manual testing facilities shall be provided for on-line testing to prove operability and to demonstrate reliability.

Plant operation should not be adversely af fected.

6. All new instrumentation shall meet the environmental and seismic requirements of IEEE-323.

7. The status of all con ainment isolation valves shall be provided in the control room and not be affected by the modifications.

8. Non-safety related radiation isolation signal will meet all of the above criteria with the following exceptions:

a. The system will not be seismically qualified,

b. Testability requirements of IEEE-279 will be met to the extent practicable.

2.1-13 1456 277

2.1.1.5.3 Design Evaluations and Systems operation In order to cover a broader spectrum of events for which contain-ment isolation is desirable, the reactor trip signal is used as a diverse containment isolation signal. Since a reactor trip signal occurs on low pressure (1800 psig) it is anticipatory of SFAS and occurs prior to SFAS initiation. Therefore the NRC directive would be fulfilled in a conservative way by the reactor trip signal rather than the SFAS signal.

The use of the RPS system would provide isolation for the follow-ing eve-ts:

a. Rod withdrawal accidents

b. Loss of coolant flow

c. Feedwater line break or loss of feedwater

d. Small steam line break accident outside containment (isola-tion of containment lines is still desirable)

e. Ejected rod accident

f. Boron dilution accident

g. Cold water addition

h. lodine spikes or crud burst after trip

i. Loss of offsite power or station blackout The 1500 psig SFAS signal would not isolate containment for items a, b, c, f, g, h and 1. Isolation on 1500 psig SFAS for items d and e would not cover a full spectrum of events.

As discussed above, lines which will be isolated on reactor trip are:

a. reactor building sump

b. RCDT gas vents and liquid discharge

c. RCS sample lines

d. containment purge lines

e. RCS letdown

f. demineralized water

g. OTSG sample lines (due to primaty to secondary leaks)

Closure of these paths by a signal that is not dependent on building pressure assures that there will be no uncontrolled radioactivity release f rom containment for design basis events.

With the exception of the letdown and the demineralized water valves , the above lines are normally isolated. :f these lines receive an isolation signal af ter a reactor trip the plant condition is not degraded. The letdown lines is normally open, and it is immediately closed by optrator action af ter reactor trip.

Special Jesign provisions will be taken with letdown line isola-tion. 11 aeither 4 psig building pressure nor high radiation exists, the operator will be able to reopen the valve on demand.

If eithe rof these signals does exist, however, the operator can only reopen the letdown valve by overriding the closure signal the valve.

}k h 27b 2.1-14

The demineralized water line is normally open to provide purging of the reactor coolant pump number 3 seal. The purging prevents boron building in the seal. Loss of this function is not a concern. Westinghouse , the pump manuf acturer, has stated that loss of seal purging has been determined not to affect the seal; in fact, at the owners discretion, some pumps are being operated without the purge water connected.

Individual high radiation signals will be used to prevent re-leases outside containment for the:

1. Reactor building sump drain

2. Reactor coolant system letdown line

3. Reactor coolant drain

4. Reactor building purge (monitor already exists)

5. Reactor coolant sample lines

6. OTSG sar.ple lines

7. Reactor coolant pump seal return (alarm only)

8. Intermediate closed cooling water (alarm only) l Intermediate closed cooling water will be isolated on high radiation in order to prevent inadvertent releases due to let-down cooler leakage into the ICCW system. Isolation of the ICCW system will not jeopardize operation of the reactor coolant pumps. The pumps can run for approximately one week with only seal water providing the cooling for the pump seals. Plant operating procedures will be revised in order to address re-initiation of ICCW cooling of the seals for periods lenger than one week.

Individual raidation isolation have been chocen in lieu of a general radiation isolation signal for the following reasons.

First, reactor trip isolation will be anticipatory of a high radiation condition. Second , individual isolation is more sensitive to isolating the source of activity. For example, a general radiation signal based on dome activity would not detect a source of activity being added to the RCDT.

Once containment isolation is completed, certain lines may have to ce reopened in order to support post trip or post accident operation. Table 3 of Appendix A provides a list of override capability for each of the lines receiving either: reactor trip, high radiation, or 4 psig or 30 psig building pressure isolation signals.

Plant procedures will govern the conditions under which any of these overrides are utilized. In genera, the prerequisite for override is a determination that neither an accident condition nor a radiation hazard exists. If eithe rof these conditions exist, then specifics as to if or when the isolation can be bypassed will be developed on a case by case basis.

Individual reactor trip override capability has not been supplied for all lines except RCS letdown. When a stable post trip} 2.1-15 Am. 5

condition is achieved, the operator can bypass the containment isolation signal at the system level in order to reestablish control of these systems. 2.1.1.5.4 References

1. Letter from Boyce Grier, of US NRC, to all owners of B&W reactors dated April 5, 1979, IE Bulletins 79-05A, 79-05B, 79-05C.

2. 10CFR50, Appendix A, General Design Criteria 55, 56, and 57.

3. B&W Company, Nuclear Power Generation Division, dated 5/22/79,

              " Recommendations for Short-Term . manges to Containment Isolation Systems as a result of the Three Mile Island Unit 2 Accident."

4. B&W Company, Nuclear Vower Generation Division, dated 5/22/79,

              " Recommendations for Long-Term Changes to be Considered to Containment 1 solation Systems."

5. U.S. Nuclear legulatory Commission. Standard Review Plan Section 6.2.4, Containment Isolation System, U.S. Nuclear Regulatory Commission.

6. U.S. Nuclear Regulatory Commission. TMI Lessons Learned Task Force Status Report and Short Term Recommendations. NUREG-0578, July 1979.

2.1.1.5.5 Safety Evaluation The selective addition of the containment isola; ion signals on high radiation, reactor trip and 30 psig building pressure does not compromise plant safety for the following reasins:

1. The system is designed as safety grade and single failure proof (except fo r high radiation isolation). Thus, the system will pc-form its safety f unction when required.

The probability of containment isolation occurring on demand is increased.

2. Spurious initiation of an isolation signal will not introduce new accidents into the plant design. Spurious initiation of any of the above signals would not isolate any components that would also be isolated by a spurious initiation of the existing 4 psig building pressure signal.

Finally, the design meets the intent of all NRC directives to Met-Ed regarding containment isolation namely the addition of isolation on high radiation, and low RCS pressure. The design meets the requirements of Standard Review Plan 6.2.4 to the extent practicable. 1456 280 2.1-16 Am. 5

2.1.1.6 Instrumentation to Detect Inadequate Core Cooling 2.1.1.6.1 System Description The purpose of this modification is to provide instrumentation for detection of inadequate core cooling as required by paragraph 2.1.3.b of NUREG 0578. It consists of the following parts: A. Connecting in-core thermocouples to plant computer. B. Providing a wide range reactor outlet (Tg) temperature measurement. C. Providing control room indication of reactor coolant satura-tion pressure margin. 2.1.1.6.2 Design Bases This modification is to provide the control room operator with information to assist in identifying inadequate core cooling conditions. High quality, control grade instrumentation shall be prov ided . To the extent practicable, sufficient redundancy shall be provided to allow surveillance of instruments by comparing dif ferent channels and to f urnish the operator with alternate information if one channel is disabled. Instrumentation shall be available af ter a loss of of fsite power. This modification shall not degrade the integrity of any safety-related system or any existing instruments which are required for safe and reliable operation of the plant. 2.1.1.6.3 System Design 2.1.1.6.3.1 In-Core Thermocouples The in-core thermocouples are presently cabled from the reactor up to electrical containment penetrations but have not been terminated at the penetrations. The existing chromel-alumel conductors will be spliced to copper wires and run to adjacent penetrations which have spare conductors. Connections will be made to the computer with copper wires. Temeprature detectors will be used to monitor the copper to chromel-alumel junctions so that compensation can be made in the computer. This method was necessary since thermocouple extension wire penetrations were not av a ilable . The method of connection is shown in Figure 2.1-8. All splices will be made inside penetration terminal boxes and will be protected by means of heat shrink tubing. All 52 of the in-core thermocouples will be brought to the computer. This will provide considerable redundancy since the operators will be able to assess post accident core conditions adequately with as few as 16 of the installed thermocouples available. 2.1.1.6.3.2 Wide Range Tg The present Reactor Outlet Temperature measurement channels (Tg) have a range of 520*-620*F. They are used for plant control. They are derived from RTD's which are of the same type 2.1-17 Am. 5 1456 281

and are located in the same Thermal wells as the safety related RTD's used in the Reactor Protection System. This modification will provide a wide range 120-920* Td outputs from the existing RTD bride without changing the range or accuracy of the existing signal to the control system. This will be done by installing a new, specially modified convertermodule across the output of the RTD bridge in parallel with tne existing output module. The new signals will be connected to the computer and will also be used as inputs to the saturation pressure instruments described in 2.1.1.6.3.3 below. This modification will be made to four Tg channels, two in each Reactor Coolant loop. A block diagram of the new arangement is shown in Figure 2.1-9. As a long term upgrade it is intended to isolate the new wide range TH signal from the existing control signals. These signals will then be seismic I and separated for use as redundant signals. 2.1.1.6.3.3 Saturation Margin Indication In order to aid the operator in detecting inadequate core cool-ing, an instrument will be provided which will display in the control room the margin between the actual primary plant tempe ra-ture (Tg) and the sature. tion temperature (Tsat) for the existing plant pressure. T sat will be computed using primary pressure measurements and compared to the wide range Tg mea-surement described in 2.1.1. 6. 3. 2 ab ov e . The temperature margin will be displayed in the control room. An alarm will be initiat- . ed if the margin falls below a pre-set value. Redundancy will be provided by computing Tsat margin independently for each R.C. loop. The lower temperature for each loop will automatically be selected for the computations. Saturation pressure margin is also computed in a similar manner so that the oeprator has the option of displaying the saturation margins in terms of pressure. The equipment used for these comput stions will be safety grade and seismically qualified. In addition, the plant computer, using the same inputs, will independently compute Psat and Psat margin for logging, trending, and alarm. A block diagram of the system is shown in Figure 2.1-10. 2.1.1.6.4 Design Evaluation 2.1.1.6.4.1 In-Core Thermocouples The copper to chromel-alumel junctions which have been created by this modification will cause offsets in the thermocouple measure-ments. However, providing temperature measurements at the junctions will enable the computer to compensate for these of f sets, preserving the accuracy of the in-core temperature readings. The splices will be protected against potential degradation by covering htem witha heat shrinkable tubing which has been qualified for use inside containment. 1456 282 2.1-18 Am. 5

2.1.1.6.4.2 Wide Range Tg Tests have been run to demonstrate that the modified converter module will give an ..eurate output over the desired range of 120-920*F. The tests also showed that addition of the new equipment will not degrade the existing narrow ri.nge 520-620*F control signal. This addition will be implemented with Bailey Controls Company type 820 hardware which has a 'aistory of re-liable operation in nuclear plants. 2,1.1.6.4.3 Saturatioa Margin Indicator The modification will provide continuous indication of T sat margin to the operator. The design will use existing reliable plant inputs. The new instrumentation will be solid state, seismically qualified, safety grade equipment. 2.1.1.6.5 Safety Evaluation None of the modifications described affects any existing safety related instrumentation or control channels. It has been con- f cluded that these modifications will not degrade any safety related systems. 1456 2813 2.1-19 Am. 5

g 2.1.1.7 Auxiliary Feedwater Modifications 2.1.1.7.1 System Description The TMI Unit #1 Emergency Feedwater System is being modified so that:

1. Both of the motor driven Auxiliary Feedwater ( AFW) pumps automatically start upon loss of both main feedwater pumps or loss of four (4) Reactor Coolant Pumps.

2. The motor driven AFW pumps are automatically loaded on the diesel generator during loss of offsite power.

3. Indication 's available in the control room of AFW flow to each steam generator.

4 Manual control independent of the Integrated Control System (ICS) is available to the operator in the control room.

5. Control room arnunciation for all auto start conditions of the AFW system is available.

2.1.1.7.2 Design Bases The TMI-l Auxiliary Feedwater System (AFW) is being modified so that a single failure will not result in the loss of auxiliary feedwater system function during a Loss of Coolant Accident. To accomplish this the requirements of NUREG-0578 Section 2.1.7a and 2.1.7b will be met. In addition, the emergency feedwater control valves are being modified such that they fail open on loss of instrument air in order to meet the single failure criteria. 2.1.1.7.3 System Design As indicated in Chapter 10 of TMI Unit #1 FSAR, the Emergency Feedwater System was designed to operate. 1) on loss of all four Reactor Coolant pumps and 2) if both main feedwater trains fail. The original system design was based on use of three auxiliary feedwater pumps. One of the three pumps is turbine driven and has a capacity of 920 gpm. The remaining two pumps are motor driven and have a capacity of 460 gpm. The three pumps are located in  ; the Intermediate Building which is designed to withstand seismic events, tornado, missiles and a hypothetical aircraf t incident. i The turbine driven pump is physically separated from the motor driven units. One of the motor driven pumps is powered f rom the class IE 4160 volt bus ID wnile the other motor driven pump is powered f rom the redundant class lE 4160 volt bus IE. The design of the ID and IE Bus has been changed so that they continue to supply power to the motor-driven pumps during loss of of f-site 1456 284 2.1-20 Am. 2

power conditions with coincident ESAS actuation. To limit voltage dip on the diesel generator during loss of off-site power  ! and coincident ESAS actuation condition, the motor driven pumps l will be loaded as a block 5 load (i.e. will be loaded 5 seconds ' after block 4 loading). Power to the turbine driven pumps is from the safety related portion of the Main Steam System. The design of this system remains unchanged and is described in Chapter 10 of the FSAR. Both of the motor driven and turbine-dri"en emergency feedwater pumps receive an auto-start signal on Icss of all four reactor coolant pumps and loss of both main feedwater pumps. This is accomplished by utilizing contacts from the Reactor Coolant Pump power monitors and by sensing the differential pressure across the main feedwater pumps. The RC pump power monitors are a safety grade system and are described in chapter 7 of the TMI-1 FSAR. The main feed pump differential pressure sensing equipment is control grade. Both of the above initiation signals and circuits are designed so that a single failure will not result in the auxiliary feedwater system not functioning. To accomplish this, the actuation system is arranged into two trains. Each train contains two differential pressure switches (one for each main feedwater pu=p), and four contacts from the RC pump power monitors (one for each pump). Power for the "A" train is from the 120 V. AC Vital Distribution Panel VBA. Panel VBA can receive power either from the "A" station battery through the 1A inverter or from the "A" diesel generator. The "B" actuation train utilizes redundant pressure switches and RC pump power pump monitors and is powered from the 120 V. A.C. Vital Distribution Panel VBB. Panel VBB can receive power f rom either the "B" station battery through the IB inverter or from the "B" diesel generator. In addition to the above actuation signals, the turbine driven pump also receives an automatic start signal f rcm the main feed pump trip circuitry. The details of this actuation signal are discussed in chapter 10 of the TMI-I FSAR. All three emergency feedwater pumps discharge into a common header. From this common header a separate six inch line de-livars water to each steam generator. Each of the two supply ' liues contains an air operated control valve (EF-V30 A/B). Under normal operations air for the control of these valves is supplied from the instrument air system. The instrument air system is described in chapter 5 of the TMI-1 FSAR. In the event the main source of instrument air is not available , a back up , source of instrument air has been provided. The back up air i supply is received from a 80 gal. reservoir which is supplied by l an 18 SCFM air compressor. Transfer to the back up air supply is j automatic and no operator action is required. The back-up air ' compressor is powered from the 1A 480V Engineered Safeguards Control Center. 2.1--21 Am. 2

To provide further assurance that emergency feedwater can be delivered when required, the failure mode of control valves EF-V30 A/B is being changed. Currently these valves fail half open on loss of control power and fail "as-is" on loss of instru-ment air. The change consists of modification to the operator such that on loss of air, the valves will fail in the open position and remain in this position. Control valves EF-V30 A/B are controlled by the Integrated Control System. The design of this sytem is described in chapter 7 of the TMI-l FSAR. Upon loss of all reactor coolant pumps, and/or both feedwater pumps, the IC positions the control valves to maintain steam generator water level. If reactor coolant pumps are available, the ICS controls are set to maintain a 30 inch water level on the start-up range level indicator. If reactor coolant pumps are not available, the ICS maintains steam gener-ator water level at 50% on the operating range level indicator. The Integrated Control System is a connrol grade s;' stem. It does, however, receive power from the Class .E power systes Specific-ally the ICS is supplied from Distribution Panal ATA. This panel can be powered from the station batteries thru inverter . A and Panel VBA or from ES Control Center IA through Panel TRA. Manual Control of the emergency feedwater control valves can be taken from the control room. When manual control is selected all active components of the ICS are bypassed except for the raise / lower voltage circuit. As further assurance that control of the emergency feedwater control valves are available to the operator, an additional manual control station is being provided for each valv e . The controls will be located in the control room and will be totally separate from the ICS. Power from the redundant por-tion of Class lE power system will be provided to the back-up controls. Manual controls of the emergency feedwater control valves can be taken f rom the control roo=. When manual control is selected all active components of the ICS are bypassed except for the raise / lower voltage circuit. As further assurance that control of the emergency feedwater control valves are available to the operator, an additional manual control station is being provided for each valve. The controls will be located in the control room and will be totally separate from the ICS. Power from the redundant portion of Class IE power system will be provided to the back-up controls. A functional diagram of the new manual controls is shown in Figure 2.1-3. A new manual loader station for each control valve will be mounted on the control board. This will allow the operator to manually set a

 +10 volt control signal into the voltage / pneumatic converter in order to control the position of the EFW control valve.      An adjacent selector switch connects the signal from the manual loader station to the voltage / pneumatic connector and disconnects from the ICS "EL" power supply to an independent 115 volt , 60 hz supply. Thus, if the EFW controls are disabled due to a failure in the ICS or failure of the "EL" power supply, the operator will have the ability to control flow to either steam generator entirely independent of the ICS.

2.1-22 Am. 3

Each of the emergency feedwater supply lines has also been provided with a flow sensing device. This device is a sonic flow device as manufactured by Controltron and will be installed downstream of the control valves before the the lines enter the containment building. The flow device is safety grade and has been seismically qualified. The output of the flow devices will transmit the signal to the main control room where meters will be installed to read flow directly. The cquipment to be installed will be safety related. Cabling will be routed as described in Section 7 of the TMI-l FSAR. The power supply for the instruments will be derived from the vital 120 V power system. Redundant Power supplies will be used. A diverse means of monitoring emergency feedwa ter flow is provid-ed by tbc steam generator level indicators. These measurements are derived from Barley type "BY" t ansmitters which, subsequent to their installation at TMI-1, have been seismically qualified and qualified for operation in a post-LOCA containment environ-ment. One start-up range and one operating range transmitter have been raised higher above the reactor building floor to avoid flooding in a post-accident situation and have had their elec-trical connections protected to prevent degradation due to moisture. The level instruments are supplied from lE on-site power sources and their wiring is run in raceways which have been analyzed to assume heat. They will withstand a seismic event. 2.1.1.7.4 System Operatian The TMI-l Auxiliary Feedwater System is a stand-by plant system which is not used during normal plant start-ups, shutdowns or o pe ra tion. The system is maintained in stand-by during plant operations and is automatically actuated upon loss of both main feedwater pumps or loss of all four RC pumps. The following table gives actuation time for the system: Ev ent Turbine-Driven Motor-Driven a) Loss of Feedwater or 1mmediate 5 Sec. Loss of RC Pumps b) Above with loss of 10 Sec off-site power (LOP) c) Above with ESAS but 15 Sec 20 Sec no LOP d) Above with ESAS and LOP 25 Sec 30 See Start-up and test data indicates that the turbine driven pump requires 18 seconds to reach full flow. The motor-driven pumps should be capable of accelerating to full speed in less than 1G seconds. Therefore under worst case conditions emergency feed-water flow should be established within approximately 40 seconds. Control of auxiliary feedwater following initiation is accom-plished by the ICS. The ICS controls the injection of auxiliary feedwater to maintain water level in each steam generator to on e of two setpoints depending on whether RC pumps are or are not 2.1-23 3 1456 287Am.

av ailable . Under forced cooling conditions, the ICS controls level to 30 inches on the start-up range since this is suf ficient to provide core cooling. However upon loss of forced RCS cooling the ICS controls steam generator level to 50% on the operating range to promote natural circulation with the Reactor Coolant System. Manual controls in the control room are available for the opera-tor to take control when needed or in the event of ICS failure. 2.1.1.7.5 Design Evaluations Table 8-1 of the TMI-l FSAR indicates that the heaviest loading on one diesel generator would result in 2513 KW or 97% of con-tinuous rating of 2600 KW. The addition of the motor-driven emergency feedwater pump will add 450 H.P to the diesel laoding or 365 KW. This will result in a total loading of 2878 KW or 96% cf the diesel's 2,000 hr rating of 3,000 KW. Since no credit has been taken for the reduction in pumping requirements following a LOCA and since the diesels 2,000 hr rating is not exceeded, the diesel operability will not be affected. A de-tailed loading study has also verified this fact and testing will be performed to further verify this fact. 2.1.1.7.6 Safety Evaluation Safety analyses performed on the 177 Fuel Assembly B&W plants have determined that the emergency feedwater systems for a 2772 Mw plant must be capable of elivering 550 gpm (total to both generators). The basis for this criteria is contained in Volume 1; Section 6 - Supplement 3 of B&W's report entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant". The analysis submitted by B&W is applicable to TMI-1. Several studies have also been performed by B&W for the 177 FA plants on loss of msin feedwater transients. These analysis ha"e demonstrated that :00 gpm or lower auxiliary feedwater flow is adequate following upset transients such as loss of power and the loss of normal feedwater flow. Therefore, the small break LOCA conditions with a 20 minute delay in auxi-liary feedwater initiation sets the minimum emergency feedwater capacity requirements. Considering that TMI-l is only a 2535 Mw, a minimum emergency feedwater capacity requirement of 550 gpm is very conservative. As discussed in paragraph 2.1.1.7.3 above, the TMI-1 emergency feedwater system is comprised of two 460 gpa capacity electric pumps and one 920 gpm capacity steam driven (turbine) pump. The addition of the motor driven pumps (automatically) to the diesel block loading sequence and the turbine-driven pump start circuit ensures that a single failure will not result in less than the minimum required pump capacity being available under all condi-tions including loss-of-off site power. That is at least two motor driven or one motor driven and the turbine pump will be available under a'.1 single failure conditions. 2.1-24 1456 288 Am. 3

The addition of the motor driven auto-start circuits and addition of these motors to the diesel block loading sequence ensures that a single failure will not result in less than the minimum requir-ed pump capacity being available under all conditions including loss-of-off site power. The TMI-l AFW design provides an emergency feed line with control provisions in line to each steam generator. The design is such tha t the required quantity of water can be provided to at least one steam generator during all single f ailure conditions involv-ing a Loss of Coolant Accident or loss of normal feed. Under steam line or feed line break conditions, when both main and auxiliary feedwater is isolated to the af fected steam generator, a single failure of the unaffected auxiliary feed line control valve will produce unacceptable results. To counteract this situation several short term design improve- nts have been implemented. A Back-up instrument air systews have been added, the failure mode of the control valves have been changed, and an additional manual control station has been added. All of these changes provide additional assurance that the IMI-l control valves will be operable when required or at least will fail in the open position. In the long term, the system will be re-designed to account for the extremely unlikely condition where a control valve sticks closed during a steam or feedline break accident. As noted above, the failure mode of the feedwater control valves, EF-V30A/B, ahve been changed from a fail-as-is to a fail open position on loss of instrument air. This failure mode is con-sidered best because it gives priority to reliability of feed-  ; water delivery for decay heat remov al . Prevention of overfill is ; a second priority and a condition which should be prevented but I without compromising decay heat remov al Several changes have been made to ensure that the operator can prevent an overfill and overcooling condiiton. These changes consist of addition of the back-up class IE powered manual control stations for EF-V30A and EF-V30B in the control room. These changes were made to back-up the existing automatic level control system and plant instrumentation systems which in them-selves are highly reliable. In addition, plant procedures are being modified to provid- guidance to the operator in recognizing overcooling incidents A .d taking prompt corrective action. The operators will be trained in the requirements of these procedures as part of the Operator Accelerated Retraining Program. These changes ensure that control of the emergency feedwater system will be available to the operator in the control room for preven-tion of steam generator overfilling conditions. 2.1-25 lg}6 2hh Am. 3

In the event that emergen:y feedwater control from the control room were not possible, operator action to prevent AFW overfill is possible based on the following:

1. The Auxiliary Feedwater ( AFW) System does not continuously operate as main feedwater does and , therefore , the oppor-tunities for AFW overfill are reduced.

2. The automatic and manual controls provided are highly re-liable and simple.

3. A large amount of redundancy is available in the instrument air system.

4. The AFW system is relatively simple so the opportunities for failure are few.

5. The AFW fill rate is slow and therefore ample time exists for ope rator action. Depending on many variables such as total flow capacity, generator pressure, prior power level, etc. , 7 to 15 minutes will be available before the water reaches the elevation of the top shroud of the steam generator.

6. Evaluations performed by B&W for another 177-FA plant in-dicate that overcooling to the above levels will not result in any unacceptable consequences for the NSSS com >onents.

7. The plant operating and emergency procedures will be modified to address the issue of steam generator overfill. These procedures will be written such that operation action in the time permitted will be assured and verified to be achievable.

As noted in the discussion above on System Design: a) The TMI-1 design provides for the automatic initiation of auxiliary feedwater. b) Subject to the limitations discussed above the design ac-counts for single failures. c) The initiating signals are powered fro: Class IE power systems. d) The A.C. motor driven pumps and valves in the auxiliary feedwater system are included in the automatic actuation of the loads to the emergency buses. 1456 290 2.1-26 Am. 3

e) The automatic initiating signals and circuits are designed so that failure will not result in the loss of manual capability to initiate AFW from the control room. f) Safety grade indication of auxiliary feedwater flow to each steam generator is being provided in the control room. This design is consistent with the existing system design (i.e., on indicator per lin is provided). g) The Flow instruments are to be powered from Class 1E power systems. Manual capability to initiate the auxiliary feedwater system from the control room has been retained and is such that a single failure in the manual circuits will not result in the loss of s." stem fun iton. In addition provisions for testing of the f aitiating circuits, although not currently included in the design, will be provided. Control room annunciation for all auto start conditions will also be provided. 2.1.1.7.7 Startup Testing and Inservice Testing / Inspection Requirements During the initial TMI-1 start-up testing, hot functional testing was performed to:

1. Verify the Integrated Control System (ICS) controls the OTSG to the minimum level set point of 30 inches during HFT heat-up.

2. Verify the ICS controls the emergency feedwater system and OTSG level for the following simulated conditions:

a. Both main feedwater pumps tripped.

b. AC hand power to the ICS lost.

c. All four RC pumps tripped.

d. All four RC pump & both main F.W. pumps tripped.

3. Verify the auto start capability of the steam driven emer-gency feedwater pumps.

4. Verify operability of the Emergency Feedwater System to supply feedater when OTSG pressure is 1015 psig.

These tests are documented in Test procedure TP 600/11. Accept-able test results were obtained and therefore no need exists to re perform the above tests. However prior to re-start of TMI-l the following test will be conducted:

1. Functional tests shall be performed to verify the emergency feed pumps start on loss of feedwater or loss of four reactor coolant pumps.

1456 291 2.1-27 Am. 3

2. A functional test shall be performed to verify the opera-bility of the diesel generators with the loading of the emergency feed pumps.

3. The failure positions for the emergency feedwater control valves shall be verified.

4. A functional test of the new manual control valve station and the auxiliary feedwater flow instrumentation will be perform-ed at cold shutdown conditions.

5. Operaability of the new back-up instrument air compressors will be demonstrated.

During the TMI-1 Startup and Test Program, tests of th e Emergency Feedwater System were conducted to demonstrate the ability to supply feedwater when the steam generator pressure was 1015 psig. The specifics of the test including test results and acceptance criteria were documented in Test Procedure TP 600/11, Emergency Feed System and OTSG Level Control Test. During a portion of the test , the emergency feedwater system was operated with the Reactor Coolant System in a hot condition at approximately 532*F at 2155 psig with no nuclear heat. The steam generator pressure was adjusted to 1020 psig and the operating main feedwater pump was tripped. The test results verified the turbine-driven emergency feedwater pump automatically started. Manual control of the emergency feedwater control valves, EF-V30A and B, was then taken and the valves were fully open to verify the pump flow capacity of 920 gpm. The turbine-driven emergency feedwater pump was stopped, steam generator pressure reduced to 900 psig, a motor-driven emergency feedwater pump started, and water level lowered to 20 inches on the startup range. The control valves were then returned to the automatic Integrated Contro] System (ICS) mode and the ability of the ICS to control to the 30 inch low level control demonstrated. The control valves were returned to manual control and steam generator water level was increased to 40 inches on the startup range. The control valves were again placed under automatic control and ICS control on low level again demonstrated. Similar testing was performed to demonstrate that during a loss of all RC pumps the ICS would control level to the high level set point. The above described sequence of testing verified the mechanical design adequacy of the TMI Unit #1 Emergency Feedwater System, verified the ability of the ICS to control water level, and demonstrated the ability of the operator to manually control the steam generator water level under real dynamic conditions. Since the mechanical design features, i.e., piping configuration, control 1456 292 2.1-18 Am. 3

valves , etc. , remain unchanged as a result of the restart modi-fications it is not considered necessary to repeat these tests for TMI-1 restart. Autostart capability of the motordriven pum ps , the ability of the new control stations to regulate EF-V30 A/B and the functicnality of the new flow indications system will be demonstrated under onditions which will not result in an additional thermal cyc1s on the steam generator auxiliary feed-water nozzles or steam generator tubes. It is not recognized that the proposed testing program does r.ot demonstrate that the operator can prevent overcooling. Ho wev er , whether overcooling w:ll occur is highly dependent en the initial plant conditions (i.e. , power level, power history, etc.) and the oeprator action taken. Therefore, a single test cannot demon-strate that overcooling will not occur and cannot adequately train the operators in manual emergency feedwater control. It is for this reason and to avoid thermal cycling of the steam generators that the emergency feedwater tests will be conducted at ccid shutdown c >nditions to demonstrate the ability of the new equipment tv operate as designed. Operator training in manual emergency feedwater flow control will be covered in the operator accelerated retraining program. 2.1.1.7.8 Instrumentation As discussed above auxiliary feedwater flow instrumentation is being provided in the design of TMI-1. Other instrumentation required for the safe control and operation of the TMI-1 AFW System, such as steam generator level instrumentation, is described in chapter 7 of the FSAR. 2.1.1.7.9 Reference Drawings C-302-081 Rev IB-0 SS-209-662 Rev IA SS-201-186 Rev 1A SS-209-663 Rev IA SS-201-187 Rev IA SS-209-664 Rev IA SS-201-168 Rev IA SS-209-665 Rev IA SS-201-169 Rev IA SS-209-666 Rev IA SS-208-203 Rev IA SS-209-667 Rev IA SS-208-205 Rev IA SS-209-755 Rev IA SS-209-031 Rev IA SS-209-756 Rev IA SS-209-032 Rev IA B-308-564 Rev IA-0 SS-209-108 Rev IA E-304-274 Rev IA-0 SS-209-590 Rev IA E-304-275 Rev IA-0 SS-208-591 Rev IA E-304-276 Rev IA-0 SS-209-660 Rev IA E-304-277 Rev IB-1 SS-209-661 Rev IA B-201-043 Sheet 1 Rev IA B-201-044 Sheet 1 Rev IA i456 293 2.1-29 Am. 3

2.1.2.1 Post Accident Monitoring 2.1.2.1.1 System Description Post accident monitoring capability will be provided in compliance with Reg. Guide 1.97, Revision 3. Pending the availability of appropriately qualified instrumentation and equipment, the following modifications can be completed by January 1, 1981. The conceptual design will be provided for NRC review by January 1,1980. Containment Pressure - Continuous containment pressure indication will be provided in the control room using a range from -5 psig to three times the design pressure of the containment. The pressure indication will be safety grade and will meet the design and qualification requirements of Reg. Guide 1.97. Red undant indication of pressure will be provided. Containment Water Level - Continuous containment water level indication shall be provided in the control room. A safety grade wide range indicator from the bottom of containment to a level of 10 feet will be installed in accordance with the requirements of Reg. Guide 1.97. In addition, a narrow range indicator from the bottom to the top of the sump with continuous indication in the control room shall be installed which meets the requirements of Reg. Guide 1.89 and is capable of being periodically tested. Containment Hydrogen Indication -Safety grade continuous indica-tion of containment hydrogen will be provided in the control room. The range of indication will be 0-10% concentration assuming commercial availability over this range. High Range Containment Radiation Monitor - A safety grade con-tainment radiation monitor for photon radiation shall be provided The withcontinuousandrecordingdisplayinthecontrolroom. range of this monitor shall be 10 R/hr and shall detect photon radiation down to 60 Kev. Testability of the radiation monitor will be provided in accordance with Reg. Guide 1.118. To our knowledge, manufacture of appropriately qualified equipment to satisfy these requirements will commence by July, 1980. High Range Ef fluent Monitor - One high range ef fluent monitor shall be installed for each normal noble gas release point. The range of these monitors shall be as follows: Undiluted Containment Exhaust - 105 uCi/cc Diluted Containment Exhaust - 104 uCi/cc Auxiliary & Fuel Handling Building Exhaust - 103 pCi/cc Condenser Of f Gas - 102 uCi/cc The design shal be seismically qualified in accordance with Reg. Guide 1.97 and the power supply shall be non-interruptible. The display shall be continuous and recording in the control room Testability will be provided in accordance with Reg. Guide 1.118. 2.1-30 A=. 4 1456 294

High Range Ef fluent Radio Iodine & Particulate Sampling and Analysis - The existing sampling system will be expanded and will include the adition of silver zeolite cartridges. The system design and operation will both decrease the activity on the cartridges so they can be handled and will decrease the xenon activity on the cartridges by the use of silver zeolite. The expanded portion of the sampling system would be placed in service following an accident and will meet safety grade criteria where equipment availability permits. 2.1.2.2 RCS Venting 2.1.2.2.1 System Description Vents will be provided for the reactor coolant system in order to ensure that natural circulation and adequate core cooling can be maintained following an accident. The vents will be located at the top of the pressurizer and at the top of both candy canes using existing penetrations. The cischarge from the vent will be directed to the reactor coolant drain tank. The reactor coolant system venting modification will be a safety grade design and will be single failure proof for both isolation and venting of the reactor coolant system. Control and position indication for the power operated vent valves will be provided in the control room. Pending the availability of the required safety grade equipment to accomplish this modification, implementation can be completed by January 1, 1981. 2.1.2.2.2 Design Evaluation Babcock & Wilcox is currently completing the analysis to justify the adequacy of the proposed vent locations and size. It is anticipated that the unique design of the Babcock & Wilcox nuclear steam supply system which allows venting the high points in the loops themselves will preclude the necessity for venting the reactor vessel head. The analysis will be provided for NRC review by January 1, 1980. Once the analysis of the adequacy of the conceptual design for RCS venting is complete, procedural guidelines will be prepared in sufficient time to train operating personnel on the proper use of the new venting system. The conceptual design plan will be confirmed by January 1, 1980, following the completion of the final generic B&W recommendations on the reactor coolant system venting modification. 2.1.2.3 Plant Shielding Review 2.1.2.3.1 System Description A design review of the plant shielding for radiation from systems outside containment will be completed by January 1, 1980. This review will be aimed at assuring access to vital equipment and assuring that vital equipment is qualified to function in the 1456 295 2.1-31 Am. 4

general area radiation levels. This review will consider those systems which may contain liquid or gaseous input from the primary system during an accident situation including the follow-ing: low pressure injection recirculation, containment spray recirculation, high pressure injection recirculation, process sampling, makeup and letdown, waste gas. Sources for each of these systems will be developed utilizing decay factors consis-tent with the time the systemw ill be operated and dilution factors consistent with accident scenarios during which the system will be operating. Field run piping will be considered. 2.1.2.3.2 Design Basis The source term to be used for shielding calculations shall be as follows: Liquid Systems: Noble Gas - 100% of core inventory Halogens - 50% of core inventory Others - 1% of core inventory Containment Air: Noble Gas - 100% of core inventory Halogens - 25% of core inventory The criteria for limiting general area radiation levels in order to assure personnel access to vital equipment will be as follcws: Areas requiring continuous occupancy - <l5 mr/hr Control Room Operation Support Center (TMI-l Health Physics) Technical Support Center (Mod / comp room and cooldown from outside control room panel) Areas requiring possible frequent access - <100 mrem /hr (Once or more per each 8 hour shift) Radiochemistry Laboratory H2 Recombiner Control Panel Liquid Waste Disposal Panel For all other areas, shielding will be provided as required to keep personnel exposures less than 10CFR20 and to maintain the integrated dose to vital equipment below that for which the equipment has been qualified. The integrated dose to vital cor ponents and equipment will be determined using the calculated radiation levels and the required length of service of each component and piece of equipment post accident. 1456 296 2.1-32 Am. 4

2.1.2.4 Post Accident Sampling Capability 2.1.2.4.1 System Description Post accident analysis of reactor coolant samples and the con-tainment atmosphere is recognized as a means to better define core damage and anticipate the need for remedial actions. The TMI-l capabilities for post accident sampling will be modified as necessary to provide key sample results on an on-line basis and to provide backup confirmatory sampling capability within 8 hours of directing that a sample be taken. The key parameters to be monitored with on-line instrumentation include containment hydrogen concentration, reacotr coolant boron concentration and letdown failed fuel monitors. The on-line hydrogen monitoring capability has been previously described in this restart report. An on-line boronmeter will be installed. The conceptual design and schedule for installation will be forwarded to the NRC by January 1,1980. The existing reactor coolant system letdown monitors will remain on scale with up to 10% failed fuel based on the FSAR definition of failed fuel. This existing monitor is deemed adequate as an indicator that significant core damage has occurred. A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed. Modifications shall be completed as necessary to ensure that personnel can obtain samples under accident conditions without incurring a radiation exposure to any individual in excess of 3 rems to the whole body and 18 3/4 rems to the extremities. The source terms to be considered shall be those previously listed under the Design Basis of Plant Shielding. In addition, a design and operational review of the radiological spectrum analysis facilities and the chemical analysis facilities will be conducted in order to identify any additional design features or shielding required to ensure that confirmatory samples can be obtained and aralyzed within the 8 hour period previously mentioned. The chemical analyses to be considered shall include both boron and chloride analysis. The results of the design and operational review and conceptual design for required modifications will be forwarded to the NRC by January 1,1980. 1456 297 I 2.1-33 Am. 4

2.1.2.5 Reactor Coolant Pump Trip on HPI 2.1.2.5.1 System Description The purpose of this proposed modification is to provide automatic trip of the Reactor Coolant Pumps when degraded primary system conditions associated with a LOCA have been detected. This will be accomplished by requiring that RCP trip be initiated when the Engineered Safeguards System has actuated Safety Injection and an increasing RC void fraction has been detected as ndicated by low RC pump motor current. The proposed logic will preclude RC pump trip during those events such as severe overcooling or very small breaks where maintenance of forced cooling is very desirable. The conceptual design described in this section is being submitted for NRC review and comment and will be implemented subject to concurrence of the NRC Staf f. 2.1.2.5.2 Design Bases Analysis has shown that a certain range of small primary breaks may result in unacceptable clad temperatures if the R.C. Pumps are tripped at a time when the R.C. System void fraction has achieved a high level. To prevent these detrimental consequences, the proposed control scheme will promptly trip the R.C. Pumps when R.C. system conditions indicate that a small break in this range may be in progress. (Until this modifica tion is in place procedures will specify operator action to manually trip the RCP 's upon actuation of Saf ety Injection). The system shall actuate when High Pressure Injection has been initiated and the R.C. System void fraction has reached a nominal value which indicates that a high void fraction may develop. It is also very desirable, although not necessary, to avoid intiation during transients such as overcooling and very small breaks where R.C. pump trip is not required so that forced R.C. System circulation can be maintained. The proposed system will meet both of these criteria. 2.1.2.5.3 System Design The R.C. Pumps will be tripped on a coincident detection of High Pressure Injection by the ESFAS and low R.C. Pumps Motor Current in at least two of the four R.C. Pumps. This means t ha t for the special case of two pumps operation, the pu=ps will be tripped on HPI alone. Redundant sensors will be used for pump current on each R.C. Pump Motor. Redundant trip signals will be derived for each motor. Electrical separation will be provided for redundant signals. No surge failure in the proposed system shall prevent a trip when required. No single f ailure in the system shall result in the trip of more than one R.C. Pump. The actuation system will be designed to be operable af ter a seismic event. However, the R.C. Pump Motor Switchgear is not seismically qualified. Provisions for on-line surveillance testing will be included. The operator will be able to restart a pump af ter trip by manual means. 1456 198 2.1-34 Am. 4

2.1.2.5.4 Design Evaluation The proposed design concept assumes that pump motor current can be used in combinaion with HP1 actuation to detect the need for an R.C. pump trip. Supporting evidence for this concept has been generated by two EPRI sponsored test programs. In 1973, Babcock & Wilcox in conjunction with the Bingham-Willamette Company, conducted a test program to investigate the single and two phase performance of a one-third scale reactor coolant pump using air-water mixtures. The result of these tests were reported under a contract with EPRI in 1977 (Ref. 1). Testing performed by CREARE under an EPRI contract using a 1/20 scale pump also shows a substantial decrease in torque at void fractions above 20%. Preliminary results of this testing were reported in the 6th Water Reactor Safety Research Information Meeting in November, 197 8 (Ref. 2) . Since torque is directly related to pump motor current, the use of pump motor current as an indicator of the fluid void fraction, based on the referenced experimental data, is appropriate. The proposed design meets the dual requirements of reliability initiating a trip when required, and not degrading plant availa-bility through inadvertent trips. 2.1.2.5.5 Safety Evaluation The system will be functional during a seismic event (except for the R.C. Pump Motor Switchgear), will be testable and will meet single f ailure criteria for actuation. No single failure will result in trip of more than one pump. Redundant circuits will be separated. Where the new system interfaces within existing safety systems, care will be taken in the design to assure that there will be no degradation of existing safety functions.

                                                                                          \k5 2.1-35         Am. 4

2.1.2.

5.6 REFERENCES

1. 1/3 Scale Air-Water Pump Program, Pump Performance Data, EPRI NP-160, Vol. 2, Oct. 1977.

2. EPRl/CREARE 1/20-SCALE TWO PHASE PUMP PERFORMANCE RESULTS, P. W. Runstadker, Jr. and W. L. Swift, CREARE Incorporated, Present at the 6th Water Reactor Safety Research Information Meeting, Nat. Bureau of Standards, Garbersburg, MD, Nov. 6-9, 1978.

1456 300 2.1-36 Am. 4 1

I TilREE Mile ISLAND UNIT NO. 1 Table 2.1-1 List of Isolation Signal Override Capability 1 solation Slj;na1 Penetration Reactor fligh 4 psig 30 psig 1500 psig Line No. Trip Radiation linilding 13uilding (SFAS) lireak Conta inment Al r Sample 108 N/A N/A I N/A N/A N/A R . il . Sump 353 C I I N/A N/A N/A RCDT 310,331 C I 1 N/A N/A N/A RCS Sampl e 328 C 1 i N/A N/A N/A R . II . Purge 336,423 C NO NO N/A N/A N/A RCS letdown 309 A I I N/A N/A N/A Demin Water 307 C. N/A C N/A N/A N/A OTSG Sample 213, 214 C I 1 N/A N/A N/A NSCCW 346, 347 N/A N/A N/A NO N/A I ICCW 302, 333, N/A I N/A NO N/A NO 334 R . II . Air Coe'.ers 431, 422 C N/A C NO C N/A R.C. Pump Seal Return 329 N/A N/A N/A NO N/A N/A (_n ley,end C = Common Signal liy pa s s ; initiating isolation condition may st ill exist.

 &                                   I      - Individual isolation signal hypass capability; procedures governing override to be developed.

A = Ant.omatic isolation signal hypass. U NO = No bypass capability; ini tiat ing condition must clear to allow reopening of valve. N/ A - Not applicable. Note: For combinat lons of loitiating signals that are all,owable, ref er to Table 1 of Appendix A.

Thart Mll.F 1%1 Ahd ttut t MQ. l l' age 1 of 3 1able 2.1-2 Ll%f or CDNTAlleu MT 15*el 4 Tire V41.Vr.5 se quignihc twmf flCAflukt Velee Valve Line Method No ras! Pos t Ac t ual Actuottee $lgoal Source te - t r i* t

• Va l ve Va l ve S l ee. of V $ l ee A. c i dent roul t ion roo t t lee N... <c r o i re Systee T ig h Tye I. Ac t ust len Fee l t lose talettog Mmt i f ied lad t ret ton Faleting Modif ied pot es l"* r an t
• l ament Alt eM t '1 . ! Sill i Al t Closed Closed t ren Tee 1.10 3.6.10 Systee puet be lesh vote t eat le tM-V2 Stil i Alt (I n .e4 Clated Tee t 't- V ) poll Air I'rce Open flaged tested at arrrerslate 1 C l os e l Tee presswee (eee RS-F).

rM-V4 Seit i Al t tTen Clac ed C l ee *4 Ice Jfl St eam t.ener st ar CA CA-V44 G lo be 1/ R IM t'pe n Closed Clam ed Tee 1,80

                                %.s=c l e                      rA-VSA ,,   C lel.e     3/8                                                                                                      1.4.1.e .10          No 56W rece==eadet taa Air          Hpen        flaeed             r[oned             Tee
  -- ,             Its          Stese ' e ne r .et ar    fA    CA-V48      C l ohe     )/9        t tMi Open        C i ne ed         fleeed             Tee        1.30                8.4.5.6.10           me.3LW r,ceemendes lee "earle FA-V58      Clabe       3/4        Air           Gren        Closed            Clased             Tee C]-} . )            W'1          l at e s ac.s l r e      IC    IC-V2       Cet e         6        EMO          Ppen        Cleged             Open/Cleoed        Tee       3.40                    4.f.8.9.10            See pese (1) lielev

@~ tactine IC-VI Cete 6 Ai r Or'n Claced tiree/Cleeed Tee L t e r is t les Le ne 7

    -M             >*f         Peel n. Water t.          (4    (A-Vl89                           Ai r C to           2                     Open        Clos ed            Cla8*

• Tee 8.50 1.5.80 i-We irl nr e..Ildlaa L b-u--._m b e9 1.et hn Li ne t a thJ DN-b2A C l o t., 2-l/2 Ftles Open e scoed C l e.e ed Tee 3.10 1.4.5.6.10 r rlf le at ion tei- V 2 6 Globe 2 - 1/ 2 t es) Orce C l ac ed C l ac ed Tee 3.10 3.4.1.6.40 f' . _ NJ! j De *I aa r a t t re r e te r- V I Cot e 2-l/2 Ai r Open Closed C l os ed Teo 1.10 1.6.10 I : _. _' ~]

g 124 Freen er laer en4 C4 C A -V ! Clabe 1/A FMo Clo.e4 Clased Cloerd Tee 1,80 f . 4.1.4.10 4 pr.sr t er f ool one

                               <**c le t f ace C A-V 2 CA-V)

Cet e 3/8 1/8 Air Closed C l ac ed C l e* *J Teg Clohe Fmi Closed Closed Closed Tee r' A-vi l Clebe 3/ A f.Mit Closed Closed Closed Tee , ,,, .c .. t .r C-o , ..t t. t. V ,, C,. . ,_

                                                                                        .                      ....        Clo...                 .a,Cl...d     T..        l. 10                 . . . . . 10        ... do.e -t .dgre..

J ro., <,43 see.r. Mu-V2. Gene 4 Air occa C a.e ed ocea/Cleoed Tee need redlet ton staael. Mt med eiere =t 3 8 be proeided l ife pe,etar Coo t ent 6pG k%-Vi Clebe 2 F)to Open Cl os ed Cleeed Tee prain T a nk 3.10 3.4.1.50 WIM:- b 4 Cele 2 Ai r Pree Cleerd Closed Tee

                              %cer til         De er t e.r %ol ent       Wie L W%-V)01     Cote          4       f.MO           Closed      Closed            Cleoed             Tee        8,50 Pr l. t enk rp                  WDI.-T )04  Cot e         4 8.4.1.50 Air            Clueed      Claned            Closed             Tee pl er t... rte lli           lat es e-!lat e          IC    I C- V4     Cete         6        Ai r           pren        Caesed            Open/ Closed       Tee                               4. F ,0.9.10 3.10                                        S&W does nat ed. trees need C ollee. Wat er to cleosif y llace se 5 st.r t y 1. 8.ie                                                                                      '

Seleele Catenary 8. Al oe see hate (I) be tene, 1 14 Int e ree.tl et e IC IC-V6 Cote 3 Ai r Open Cle.eed Cleg ed Tee 3.10 ' .4. F. 9.10 See Note (1) below. Coolles to t tiet Caeltag rolle a w y - - i e e g rm - m s a r j, . j lc - Ch V4 N

e n.., r us a t sm. ,,n , t,.t. i T T?,le 2,1-2 (Cont'd.) l_IST nF CDNIAINtTNT 15Dt.ATit'ft vAI.vf S pgflR ahC tw'Dit f rATEUNS valee velee Line Method Motent poe t Ac t ua l Actuatige Signal Sanece l'e a t rat ' nt ve t ee valec Sl re , of Velve Ace lslent Pnettinn Pagi t ing t- . te r e lc e Sy s t ee Tag h Type la Actuot tne Poeltton e' n t at i ng #Ioda f ted Ind ic ot t en histing Mad t fled Notee , im p. m. t or p l e di an Aos apt- v l 4 ave t er. As Air closed Closed Clee ed th.t le t twere tee 8,:* 8.4.3.10 fly L l ae An -v l a autter- es out closed Closed Cleved fee fly Isa pe er t ar raal. int Ns ps-vil Car e A EM Open Closed open/Cleoed Tee 8,80 See pote (1) below F,9,9,10 ruer not.e rooll m S ie ce Sntele .

            .P           pa ir v ar t aaleat           N5                  Cate                                    opes 115- v 4                    8         1390                     Cleevd            Oren/ Closed     Tee        3,80                    7,0.9,40   Se e Not e ( 3 ) be lee I~aar '%t nr                           N5-v))     Cate             9         flot         Open        Cleted            Dyee/ Closed     Ice roseline Ed st r e pc t we en is t            pr e. t ar nu t id tag        uL       wt.-v514   Cate             6         Al t         Closed      closed           Closed            Tee        1,10                    1,4,5,10   stw does not addreee 5.e p lir e l e                        WL-V 511   Cate             $         Ai r         Cleoed      Cleaed           Closed            Tee                                           need on radiat ten algaal alt             pr ar t er Pie t i d i ng      B3       be-V24     Cate             S    F.MG           pree           Closed             Dree             Tee        0,10                     7,8,10;  Betale e pela etenal f.n r e . i Air                                                                                                                                                           er p 2.10;    onlese eetle end pleine rea l e r e %pply                                                                                                                                                         or 4,5,10     ins te, a.s, ere made li ne                                                                                                         'o                                                                        Seleste Categasy I 4/7             peactor Rutiding               ES       RS-47      Cate                      Air           Open        Cleeed           Open              Tee                                 7,0,103 O                                                                                3.10                               setele 4 pela etenal horeil Ai r                                                                                                                                                               er 1,2,80;    ealese cell * *ad pipt as r oo l e e s petwee er 8.1,80     Inolde R.9, are made

n. l ac Selente Cateneey &

        *19             Brac t or Auf tding            All      AH-VIC     But t e r-     44         DFt           Clamed      Closed           Closed            Tee        1,4,, *h                 8,4,5,10 f alet Porre                                      fly l i ne                                  All- f l 0 Rotter-       48          Air           Clemed      Closed           Cleo ed           Tee Ely vitse Ar q uet Inn Si ge,1 Source
     ) 4 ruta ve st t or buildlar preneure l e nl at len                      7) Claestly line to Selsele Cat egory i

2) lis* re lg ('et As ) g en t at ton 4) 30 pelg remeter besilding pressere lentation

  )) e s f t et t on a l er=,      ..t er etne act ina required               9) Llee break lealet ten elgesel er protect f ree pipe eektp and jet lepingeoest

4) High f.edlet tose (pan- ealet y) f eeletion 80) Feente menest centrol

5) Peactor e rlp t oe l a t lan A)

• eve r t i de rapaht li t y na In..'etdwel ce leen M

hotrst (I P !.cc esplanat iese la t est o f T UR - No, TMI-137 ps.10, para IV 3) e) It) and all) regarding line break t ee t at ion. A B lac brent l en t a t l an le not regelred provided the line can withetend, or le protected f ree, jet loplageoemt and the only pipe whtp t hat can Iseals it to the R. C, piptag, M U1 b

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                                                                        -v44 on the sampling line downstream of             nu Git
                                                                        -V5B           the containment isolation valve and            (New) upst ream of connection for Turb.

Plant sampling 309 I.ctdown 1.ine to Hip HU-V2A Purea; ration

                                                                       -V2B Utiltre existing Rad. Honttor RH/L-t           Inline Deelne alizers                                                  located outside R.B.                           (fix i s t ing) 32R rressurtrer and               CA             CA-VI Reartor Coolant                                 -V2             1. orate the monitor outside the R.R.         Strap Sample i.ines                                                    between t he isolation valve an.l the         on GH
                                                                      -V3             sample cooler.
                                                                      -Vil                                                          (New) 329         Rear tor Coolant              PRI            (9J-V 3 %A Pumps Seal                                                      1.ncate the online radfa'lon monitor          Strap
                                                                        -335 Return                                                          downstream of the containment isola-          on Ull
                                                                       - 3 3C
                                                                       -33D          tion valves outside of the R. R. for          (New)

Afarm Operator action is required to close valves. 330 Reactor Coolant WIM; WDG-V3 tirain Tank locate the monitor on the outside of Area Monit or, and vent the tank. 334 -V4 strap on GH Reactor Coulant WUI, WDt,-V 30 3

1) rain Tank (New)

Pump 1)lscharge -V304 3)b Reactor Building AH All-V l A outlet and

                                                                    -VLR Uttitre the existing purge outlet              Inline nnd          Inlet Purge 423                                                           -Vic t.i ne s
                                                                   -Vln            line Ra<f. Monitor RH/A-9 located              (Existing) outside of R.R.

" 353 Reactor Building Wu t. WUL-V5 34 .@a. Sumt. tira i n incate a Itquid radiation monitor Sump I.lquid

                                                                     -V535        in the R.R. Sump

@ 102 - Honttor Intenmediate Cooling IC IC-V2.1 (New) 333 Supply & Return tocate the radiation monitor on the a nit -V4,6 Strap 6" IC return line between valve on Ctl W IM IC-V3 and the 2" pump rect rc. line. (New) CD 4

BYPASS

                                                                            "      "   'F PRESS. SWITCilES ON                                                 ---

SIAIN TilRBINE CONTI OI. ')II, H I STA B1.E 4 POWFR I.EVEl. N.I. D C 11 A PS PS PS PS l Il ( 1 . - _ OTilER TIIREE _, CliANNEI.S SIMil.AR CONTACT ITUFFER "IURRINE TRIP lI D- TRIP SICNAl. PRESS. SWITCilES ON F.W. PIIMP TURBINE CONTROL. OII. D C 15 A PS PS PS PS _ CONTACT

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PS PS PS PS t j BYPASS OTilER TilREE A CllANNEl.S SIMII.AR LD CilANNEl. "A" RPS CABINET FIGURE 2.1-1 SAFETY GRADE RIMCTOR TRIP ON TURBINE TRIP OR F.W. PUMP TRIP

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                                                         "     SIIPPLY                                                                              4--
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APPENDIX 2A To Be Submitted Later, 1456 315

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3.0 PROCEDURAL MODIFICATIONS 3.1 GENERAL The preparation, review, approval and distribution of procedures at Three Mile Island is accomplished in accordance with the requirements given in Technical Specification Section 6 and Administrative Procedure 2001 "TMI Document Control". The TMI Technical Specification establishes a Plant Operations Review Committee (PORC) and the requirement that PORC review each nuclear safety related procedure and administrative policy. The PORC signs Nuclear Safety Related Procedures and recomends approval by the Unit Superintendent. Administrative Procedure 1001 establishes the format, content, review and approval requirement of all procedures. AP 1001 further establishes the requirement that all procedures relating to nuclear safety be reviewad every two years and defines the mechanism for that review. The PORC is composed of an inter-disciplinary team of engineers and plant technicians who advise the Unit Superintendent on all matters related to nuclear safety. Section 6 of the Technical Specifications defines the composition of PORC as follows:

a. Unit Superintendent

b. Supervisor of Operations

c. Supervisor of Maintenance

d. Unit Electrical Engineer

e. Unit Mechanical Engineer

f. Unit Nuclear Engineer

g. Unit Instrument and Control Engineer

h. Supervisor of Radiation Protection and Chemistry

i. PORC Chairman (Unit Superintendent-Technical Support)

j. Station Engineers assigned by the Unit Superintendent A Quorum consists of four members, at least one of whom shall be either the Chairman or Vice-Chairman of the comittee and the quorum is limited to no more than one alternate.

The Radiation Protection and Chemistry Supervisor or some other individual knowledgeable in Health Physics shall be present at PORC meetings whenever Health Physics related procedures or policies are reviewed. To accomplish the review and revision of TMI-Unit 1 procedures they were divided into two groups. The first group (Table 3.1-1) is required to be reviewed prior to restart of Unit 1 and the second group (Table 3.1-2) will be reviewed in a timely manner not necessarily prior to restart. As reference material to accomplish this review the following sources were used:

1. B & W recommendations

2. NuReg 0560 Staff report on the Generic Assessment of Feedwater Transients in Pressurized Water Reactor Designed by Babcock &

Wilcox Company i B6 317 3-1 Am. 5

3. i'uReg 0576 TMI-2 Lessons Learned Task Force status report and s1 ort term recommendations.

4. I&J Bulletins 79-05, 05A, 05B, 05C.

5. ACRS Interim Report #3.

6. Order and Notice of Hearing dated August 9, 1979.

7. ACRS Recommendations.

In addition to the above documents TMI-Unit l's emergency pro-cedures were used at the B&W Simulator by the operating staff during training exercises. Information and recommendations from these training sessions were fed back as revisions to procedures. This procedure review was started in early May, even though it was recognized that a need for further revision would occur as systems were modiCied in preparation for restart of Unit I and as recommendations changed. The review and revision of Group 1 procedures (required prior to startup) will be completed before startup. The Group 2 procedure review startsa in August and is anticipated to be completed in 1980. 3.1.1 Emergency Procedures The Ecergency Procedures have been and are being revised to include the following:

1. An objective statement was added to the follow-up action.

2. Incorporation of the philosophy of re-checking key parameters using alternative indicators where alternives are available.

3. Incorporation of the philosophy of using multiple plant parameters to judge reactor coolant conditions (I6E Bulletin 79-05A, Item 4d).

4 Stressing the heat transfer aspect of maintaining adequate core cooling at all times (I&E Bulletin 79-05A, Item 3 and NuReg 0578, Item 2.1.9b).

5. Incorporation of NRC Bulletin guidance including adequate sub-cooling, immediate trip of RC Pumps, non-defeat of E.S.

Equipment unless continued operation results in unsafe plant conditions and recognition and prevention of void formation.

6. Incorporation of the lessons learned task force's recommenda-tion on operator performance during small break loss of coolant accident; improving operator recognition and response to conditions of inadequate core cooling.

3.1.2 Administrative Procedures The Administrative Procedures have been and are being revised to include the following: 3- 2 1456 318

1. Formalizing shift relief procedures through the use of turnover checklists; requiring signatures of both oncoming and offgoing shifts and listing safety related systems removed from or returned to service (I&E Bulletin 79-05A, Item 10 and NuReg 0578, Item 2.2.lc).

2. Incorporation into surveillance procedures major valve and switch position checks ot alternate trains of emergency equipment prior to performance of surveillance testing (I&E Bulletin 79-05A, Item 10).

3. Assurance that surveillance procedures require a specific signed switch and/or valve alignment steps to be used to restore emergency systems to service (I&E Bulletin 79-05A).

4 Verification by inspection of the operability of redundant safety related systems prior to removal of any safety related systems from service for maintenance or surveillances and fication by inspection of the operability prior to return to service af ter testing. 3.1.3 Surveillance / Preventative Maintenance / Corrective Maintenance Procedure These procedures have been and are being revised to include the following:

1. Assurance that no more than one (1) safety train is defeated during maintenance or surveillance testing (I&E Bulletin 79-05A, Item 10).

2. Incorporation into surveillance procedures major valve and switch position checks of alternate trains of emergency equipment prior to performance of surveillance testing (I&E Bulletin 79-05A, Item 10).

3. Assurance that surveillance procedures require a specific signed switch and/or valve alignment steps to be used to restore emergency systems to service (I&E Bulletin 79-05A).

4 Verification by inspection of the operability of redundant safety related systems prior to removal of any safety related systems from service for maintenance or surveillances and verification by inspection of the operability prior to return to service af ter testing. 3.1.4 Operating Procedures The Operating Procedures have been and are being revised to include:

1. Changes necessary to conform to plant modifications

2. Incorporation of a natural Circulation Procedure (I&E Bulle-tin 79-053, Item 1) 56 519 3-3

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1456 321

bHOUP 2 "^'CEDURES TABLE 3.1-2

Nvt00PI fin.                   fliti                                                     PPDCIDURE fi0.            IlllE 1001                  Do(i w nt fontrol 1101-1         Plant limits and Precautions 1003 Radiation Protection llanual                                          1101-2         Plant Setroints 1005                  Security Pror.edures/ Plan 1101-2.1       tiadiation finnitoring System Sctroints 1006                  lill Retraining Prograni 110:-2.2        Iransient Cimbustible List 1001                  fontial of Records 1101-3         Contalrunent Int. A Ar r ess I imits 1008                  f. cod Hou .ri reping 1103-4 Ralance Of Plant 5ctenints 1011                  fontrolled Fey Locker fontrol 1102-12        liydingon A.I,litian 4 Degad fication 1010.                 tjualit y h.ntial War clam'.in9                                       1102-13        Decay i; cat Rmoval ny Olst; 1019                  flisillr icalinn of I'ersonnel Per formimj 5 pee.ial Prncedures       1102-14 Itcarlor Bldg. Purging & Ventitur 10/0                  Cleanline;s Pcquisewnts 1102-15 Iill & Drain f uel Iransfer I: anal 1021                  Plant ikwfifications 1103-4          Soluble Poison Concentration Control 1022                 Control of fleasuring lest EqulPment 1103 15        Reactivity palance 1023                  Test f qulgent Recall 1103-15        llcat palance Ca culatinns 1024 fontrol of Ill! Q. C. Records                                         1104 1         (n,, y y oga ,,y g ,

1025 Special fluclear flaterial Accountability 1104-3 Condensate Clusic al iced 1030 Control of Access lo Primary Systept Openings 1032 Illssewation of it f nmation "" ## " 1031 Operating ibwo's an<l Standing Orders " 1034 Control of Crmbustible flaterials (Unit #2 Only) 1035 Cont rol of Iransient Crebustilile Materials I "9 I" # s ,,

                                                                                                                  ,, g a! )

1 0:13 p:= Administealite initrols - Fire Prot. Proo. Plan 1104-11 Nuclear '.civice Closed Cooling Witer systm 1042 LD

              & linit f2 til Systm, I ist and Priest Req. Unit #2 Only                     1104-12 Secnenfaej ',ci vice s Cloud fooling Idaler system 1043                 E ngineering Change flodifications                                                                                 '

N N

GRUUI' 2 ' 'CEDURES _ TABLE 3.1-2 EP0ggRE_tio,. i_ lit E, rRoCEOURE NO. 1111 E 1101-13 Decay lleat Closed Cycle Cooling System lin4-24?t IIO4-14A Intenne.Ilate Didq. (Inc. Imer. IW he,p Area) Sica ,f.en Compartment System 1104-241 Tur bine Hldg. A Ifealer Pay 1104-14B Operating Floor ventilation System 1104-24J 5ervice uater Post tooling Tower hm,p finuse 1104-14C Reactor Compartment System IIO1-24L Sewage hmipliw; it A V 1104-140 Rn Recirculation System IlO4-23H Diesel f.cncratur plag, ll04-14E Industrial fooler System 1101-25 insin unent A fnnt s ol Air System Il01-14F Reactor fluilding linit Ilcater System 1104-26 Hitrogen supply Sy,trm IlO4-15A Aux ?. fuel llanilling Dldg. Sup & Inhaust System 1104-26A Hittogen Blanleting Icadwater lleaters 1104-15B spent Fuct Pump Area 1104-27 h te pt sposa y . g,.o,g 1104-15C ttucicar Service Closed Cooling A Decay lleat Pune Asca 1104-28 Padaging A Solid. of Solid A Liquid I!ntvaste 1104-16 1104-280 Penetration Cooling System Solid Ra.1 Wista nis, sys, r,,patg gnq p nggg, attive m en 1104-18 flakeup Dcmineralizer Neutralizing Tank Disch. Iloj.yac p, g , t,Isposal of Dewatere d Resin ind Pre (nat 1101-19 f ont e nt Dullding Ventilation System U ligulil Waste Disposal System 1104-20 IluiI HInck system 1104-29A Reactor foolant Cleanup Proc. 1101-21 lenotration Preu uripallon System 1104-298 Reactor Coolant Evap. Processes 1104-22 Cycle flake-lip Pretrealment 1101-29C Spent Fuel Cleanup Process 1104-23 Cycle 34 4 IP-UP DCSI""'8II'"F 1104-290 Decay llcat Cleanup Processes 1104-24A Cycle flakeup Water Pretreatment flouse 1104-29E Bleed A feed Processes 1102-24R Intake Screen & hanp llouse Ventilation H04-29H Transfers, Cleanup and Ivap. of Water f rom Hisc. Haste Storage Tank 1101-24C River Water Intake Chlorin. Ilouse IIM-293 RC Drain Tank Iransfers s 1104-24D Service Dullding IIO4-29K f ransfers. Cleanup and tvap. or Water f o om IIO4-24E Circ. Water Pump linuse laundry hte Storage tank LD

@   1104-24F       Circ. Water Chlorinator flouse                                  llM-29L           Neutrallring Processes W  H04-24r,       Substation Relay Control flouse                                 1104-29(1 Neutralliing Waste fican-up and Ivar.

N 1101-29N us Tr ansfer s, Cleanup aval Ivap of liccant f rm 5 pent Rc;In A thed Precoat stoiat:c l. inks s

unuul' 2 I" L t.UUHL5 TABLE 3.1-2 rR0rr00RE NO. IlllE PR9CEDl1RE NO. Tille 1104-29P stae top an<1 encrat ion of the RL Ivapoiator lima kiern ou e Vrntilation [ qui eent t R.W. 1104-299 Startop ar.d Opetation of the Hist. Wasic tvap. I30b49 An"lliary System Operating Procedur e - IIO4-29R (hncstic Water cont ent. Was te ".torage Tank t vap. 1104-295 1101-50A 1:ansfers Frun Waste Ivap (ondensate Storaye IndustrialiWaste Treatment System T anh 1104-500 Industrial Waste filter Systnit 1101-291 Rad Waste Iransfers rein Dnit #2 to Unit il ' " " " " 11mi-290 fint in use - !~nr Ininemation Only 1101-52 resin Regenerat ion an<l Reptarrment for 1104-29W tilsc. Trans fers ppg, orating Demin. 1101-29Y 1104-53 Precoat Filter Osmratten Resin Replu e. for Waste tvap. Cond. 1104-30 Nucicar River Water 1104-51 l eading anil f lustning flakeup f. Purif. 1104-31 Seconilary Service River Watei ikwinerallicrs 1104-32 1104-55 Decay !! cat River Walt r System 11 car. tor D1dg. Alnusphere Cleanup System 1104-13 Scoren llouse (quip 1104-56 (07rire T.xtinguishing Systen 1101-31 lue hine 'lli Constitioner and Supply 1104-57 fuel oil Storage and Transfer System 1104-35 Circulating Lter (hlorination & Chemical 1104-50 Arfdit ion Systna e,cwage lift Sy<.tew pt) 1104-l,0 1104-36 Pe rercration rhemical Cicaning Itasin River Water Chlorination Systen: 1105-8 1104-31 Hechanical Draft Cooling Tower Radiation flimitoring System 1104-38 1105-9 Contial Rod Drive System Reactor Illdy. Imerg. Cool. River Water System 1104-40 1105-10 C'm'pu t er Plant Suny & Dialnage System 1104-47 110' 11 Sec. I'lant & Aux. Sys. - NNI inst.

                   'itation Services Air 1104 43                                                                                 1105-12           Copamenications System fluclear Plant 5ampling 1101-44      A                                                                          1105-13           Security Systews luehine Plant Sampling CD                                                                         1105-14 1104-45       &    I lee l'rntection Systrw                                                               t oose Parts fking tor System 1101-46                                                                                 1105-15            transient tionitor I. lect ric el:>at iracing Il04-4/A       N                                                                        1105-16           til 11and Padin System Rerialmett Water System C>

1101-418 themie.a t A.I.lition rheclear

u wol> 2 F tLUURES TABLE 3.1-2 PROCf DURE NO. IlitE PR(KEDURE NO. f til,F. 1106-4 Auxiliary Unilers 12b2-2 Sta filariout A Station Hlar60ut With loss Of Both Dic'.el Generators 1106-5 luiteine foypass 1202-8 1106-7 Stator rnoIIng System reti lenin f ailui es - (Po llalf unc t ion Act ion 1106-R ltyitrogen Scil Oil an 1 Gas System. " '"'"^' 1106-9 1202-17 loss of intrimliate fooling Sy-tem lorhine tube Oil runp System 1106-10 Turbina Gland Steam Supply system 1106-11 1202-31 Iire Isolated Phase Bus Duct Cooling 1202-32 FInwl 1106-12 Ixtraction Steam, lleater Vents & Drains 1202-311 lij6-13 romfex System Nuclear % rvice River hter railure 1106-14 1201-5 tilqfi ration Ennfuctivity In the Con.fensate liain Stram anit/or ice <lwater Systra 1106-15 Flain A Auxiliary Vacutan system 1203-7 tian.1 t ais . f or r)na. rower litt A core 1106-16 Powce tei.l a l . OISG 5ccon<fary F 111 Orain A Layup 1106-17 Turbine liigh Pressure F luid 1203-10 ffnanticipate<i Criticality 1106-19 1203-16 Auxillary 5 team f ross Connection ter. hr, an,i sintor it.ilfunction 1101-4 '" ilect rtral Dis trit ution Panel I isting 1101-5 Eles trical Distritiution C<waponent Listing " 1203-21 5.5.0.C. Sys tren Iallure 1203-281 rost An i. tent 112 I"'4" l203-34 l'onteol 1:ulleling Ventilatinn 5y.tre 1203-40 Vibration A t oose Parts Honitor System Cr1 Ch u N LT1

unuul' t l'""tLOURLS TABLE 3.1-2_ PROC [ DURE NO. 18111 1300-IAD fontinuous Wet fluorescent fiagnetic Particle 1300-IV6 Inservice Inspec tion of Raattor Vouct Procedures for l'e cnute Retaining Bolting Closure llead Stu.ls. Nuts anit Washers 2" Dianieter and Greater 13n0-IV1 Inservle e inspect ion of Reacto Vessel 1300-1AG lit ter.onic Examination of Crian Pressure Pressure Retaining Colting Below 2" Itousing Welds Dia, t'ontrol Rod prive Gasketc.I loint 1300-1D Dolting Ultrasonic Examination of Reactor flange to Vessel Weld and flange Ligaments 1300-IV8 Inservice inspection of Reactor Vessel 1300-lE Closure IIcad fladding Hanual txam of Piping Butt Welds A Long Welds 1300-IVIO Inservice inspection of Reactor Vessel 1300-1F an1 Intcenal Structures liltrasonic Isavalnation of Reat. tor Closure 11 cad to Flange Welds 1300-IVil inservite Inspection of h rssurlier 1300-lG Clmanicirntial and logitudinal Se,un Welds Automat ic illt ra .onic [namination of Reactor Closuic Head Crum floiz1c and Penetration 1300-IV13 Wolds Inse vit e incrert ion of Pictsue lier llcaler ftundles and flanway Polling 1300-1H Ultrasonic fxamination of riping and 1300-IVI4 Inecrvl<c Inspection of recnurlier florzle ill-Hetallic Welds rotting Below 2" 111a, Relief and Safety 1300-1J Valve Dolting 111trasonic lxamination of Preuurizer VencI Wclds 1300-IV15 Iv.crvice Inspe<.tlon of Perssur iscr 1300-lK Integially Walded Vcssel Suiports Ultr asonir [xaminat ion of Sicam Genes tf or I!pper and trwr 11 cad to Tubcsheet Welus 1300-IV16 Iriscrvice Insrcrt ion nf h cuurtrer an.1 Suppm t structure Welds Vessel cladding 1300-IV11 Inservice inspection of Steam Generator th q.It flanual Hitrasonic f ram Proc for Stram Gen Norrie inner RailOJs Tubcsheet to ficad Weld 1300-lQ 1300-IV1B Inservice Inspection of Steam Genarator I!ltrasonic lxamin-stion of IdsselIrd I lyv; heels on Reactor Conlant hanp Hotor inlet and Outlet Norzles Rotors 1300-IVI9 Inservice Inspection of Steam Cencrator 1300-li flanway Golting fianual liltrasonic [xam. Prnc for Pressure Retaining Bolting (Rolts, studs & Nuts) 1300-IV20 Inse vice inspection of Steam Generator 1300-1U (11trasonic Examination of Raactor vessel inspection Opening Bolting (Dolting Closure Head Pressure Retaining Studs tielow 2" Dlal A and Nuts g 1300-1V21 Inservice inspcrtion of Steam Generator & 1300-1V3 Inservice Inspection of Reactor Vessel Integrally Weldad Vessel Supports Control Rod firive Norzle and Pressure 1300-IV?2 Inservice Inspection of Steam Generator L rJ Housing Welds Vessel t:ladding IV

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GROUP 2 1CEDURLS _TA_BLE 3.1-2 PROClioRE NO. Tlit r PRottiURE fio. Il it i. 1303-5.1 R.R. Cnoling & Isolatn 'iys. toqic 1303-11.24 Rit local leal. age Penetrat ion Pre sur-(hannni 1. Comp lost tration 1301-5.2 t oadiry .c.prnre 4 Somponent Test A IIPI logic f.hannel 1303-5.4 1303-11.26 frergency Icedwater lumps R.n. Isolatinn Valve rycle Test Required 1301-5.5 1303-12.3 Control Room Tmcr9ency Filtering Venting ni HO hers and tir I Lines Systrm (OP. Icsts) I'"1"coryactueling Intenal 1303-6.1 1303-12.4 R.II. Integrated leakage Rate Test Vent ing of TWI hwps arn! IP I l. ines 1303-6.2 flyrfrogen Purge Operating Test 1303-12.13 1301-8.I I lic sy . tem 41u'.h at 2* Drains-Deluqe/ Reactor Coolant System Test sprinkler %yst. 1303-8.4 1303-12.14 Reactor Duilding Spray System Ilec Protection Instomientation tion-1303-9.9 Functfonal Test Ing of Ifydraulic Snut,bers 1303-12.16 1303-10.1 II e System Testing Air Tunnei Deinqa P.D. Purge System lunct. Test 1303-11 1 1303-12.11 fontrol Rods (Drep Times) five system icstIng Hiscellancous peluge 1303-11.4 funct. lest flefuoling System Interlocks 1303 4 2.18 Iler sye Nonle Mak Int 1303-11.6 Spent fuel tooling System 1305 1.1 1303-il.fi Heather Station Dilly Checks Requireil liigh and inw Pressure Injection Interval Dally 1303-11.9 V301-2 R.B. Emergency Cooling System Coric Acid Acid Tank Hix Tank OR t-laimed Doric 1303-11.10 L5 Test 5)strm Imciq. Loading serl. A PWR Trans. 1303-11.16 1301-3E Occay Heat Reimval System leakage Reactoc Coolant System (radiochemical 1303-11.18 R.it. local l eat Rate icst ings '0' U# 1303-11.19 hn t. inn Omrspeed Testing 1301-11.20 Secon<lary Coolant Activity Reae f or Itallding Acress llatch Interlocks f e cq. 6 rio. 1301-4.8 Z> l'rimary Conlant Isotopic lodine Analysis LD 1303-11.21 1301-5.6 & (ort Flooiling system Valve opriability Cove Flooding lank llater Sarnple lest 1301-5.7 Spent Fuct Pool u 13')1-11.22 flain Steam isolation Valves a Hy ' m e Ian Concentration CD 1301-6.6 Lodium Ihiosulfate Tank roncentration 1301-9.10 River Water Disciarge Sampling

GROUP 2, ;0CEDURES TABLE 3.1-2 PROCEl>URE NO. TiltE FRottl4*E flu. IIIL E 1301-4.6 Station Storaun Catteries-Required Interval Weekly 1302-5.12 Pressurizer Tone. A tevel Channels 1301-5.8 Station Batteries 1302-5.13 Cmts al Rod AMolute resition 1301-6.2 Strong flotion Accelerirrieter Battery Checks 1 4 .14 rnntiol r,%1 relative Position 1303-11.11 1302-5.U, Station natterles (toad Test) rnic r ien,$ (ants, r,cssure A l evel thannels 1303-12.11 A/B Halon System Pres, and Weight Checks 1 0 -5.17 Hancup f ank inci Channels 1301-R.1 R.B. Annual Inspection

• 1301-9.1 R.R. Structure Integrity Tendon Surv. 1302-5.19 leniatad water storaqa Tans level Prograat In<ticator 1302-5.20 D A H I level & y,wp tha,,n,1 1301-9.12 Sulfate Ion Accountability 1302-5.21 1301-8.2 Diesel Generator, Annual Inspection RectatemlIwric Teno Channel A id tilx Iank level A 1301-10.1 Internal Vent Valve Inspect. & Exercise 1302-5.22 fontainnent Temperature 1303-11.2 Pressurizer Code Safety Valves Setpoint 1302-5.24 invironmental Plonitors Cal.

Verification 1302-5.25 R.D. %nnp tevel 1303-11.3 Hain Steam Safety Valves 1302-5.26 orsn level channel Calibrat ion 1302-3.1 f7uarterly Calibration Radiation tioni- 1302-5.27 toring Systems Sodium lhlosultate Tank Icvel 1302-3.2 1302-5.28 Sodlinn Hydroxfite Tank level Ind. Strong Motion Accelerenieter 1302-5.32 1302-5.1 A 5.5 R.C. Temp Channels A Pressure Temp Weste cas Ca rnessor Pressure Comparator 1302-6 1302-5.2 A 5.3 PPS liigh and Low RC Pressure Channel Calli of tk.n leth S toc Instr lised Calibration 1302-14.1 1302-5.4 RC Flux riow Calitiration of Inservice Insp. Related instemients 1302-5.6 Pump - Flux Comparator D' 1302-5.7 liigh R.D. Pressure Channel 1303-4.11 " liigfi an.1 t em Pressue n injec tion Analog 1302-5.8 t harmels A H.P. A 1.P. Injection Analog Channels 1303-4.13 R.R. Imcrern(y timling A Isolatien w 1302-5.10 R.D. 4 PSIG Channels System Analnq tiene+1s 1302-5.11 R.D. 30 PSIG Pressure Channels a la inn n. S W sws g FN) 1303-7.1 Intremeiliale Range t hannel 1303-7.2 Sone e r Hinv Chann, l

bKUdi' 2 l' ;EUUR_ES_ TABLE 3.1-2 t'ROCE0tmE NO. I lit.E PROCT 0trRE N0. Illit 1303-11.23 l'eac tor DIdg. Local teakage-Fluid Block System 3301-til Fire System V.seve t.ineup Verification 1303-12.5 3301-It2 tmergency Plant Radiation Instrinnentation CO2 Fire rootectior. System Test Check 1303-12.8A Fire Protect Instr. Funct Test Cont 3301-02 Ridg Elev 355' Specific Gravity Check-Diesel Fire riarps 1303-12.8B 3301-R1 fire service Diesel Ecgine Inspection fire trotect Instr. Funct Test Cont B1 rig Elev 338' 3301-WI fire System Water Source level Cherk 1303-12.8C rire Protect Instr. Fur.ct Test Cont 3301-W2 fire System Diesel Dattery Chect Illdg T lev 322' 1303-12.80 3302-R1 twrg Plant I%) Instr Calib Reqd. Inter-ris e Protect Inst r. runct Test Refueling int Dietel Generators 1303-12.IIE 3302-sal fletcorological Instrtsnentation Calibrat ion Fire riotect Instr. Funct icst Screen flouse 1301-12.8r Fire Protect Instr. Innct icst Aus A 3303- A2 fuel llandi Bldg. f lee System Itain llea<ler Flush and loop fest 1305-1.2 3103-ft1 Fire rest Periodic Operation Win <f Sreed and Wind Disection Calibration 1305-I.3 3303-Q1 Fire TNair Diess.1 fuel Vertical Ir v rature Calibration 1302-5.30 3303-R I rire rimp 5 tart Circut D6esel Generator Protective Relaying 3303-P2 Ilre rui.P Capacity lesting 1302-5.31 4160 V. D & E BUS Unelervoltage Relay System 3301-3V I Fire systeri r. pability rest 3325-5Al (t><wilcal itelease inventne y 3391-5Al f tre D>.lsant inspect ton 310fi-l tin'.i f1 & Unit #2 fondensate Cross-Connect U C i ( (

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4.0 EMERCENCY PLAN FOR THREE MILE ISLAND NUCLEAR STATION UNIT 1 See report following. 1456 333 4-1 h. 4

TABLE OF CONTENTS Section Topic Page 4.1 DEFINITIONS 4-8 4.2 APPLICAEILITY AND SCOPE 4-16 4.2.1 Applicability of the Emergency Plan 4-16 4.2.1.1 General Information and Site Description 4-16 4.2.1.2 Restricted Area, Exclusion Area, and Low Population Zone 4-19 4.2.1.3 Population and Population Distribution 4 20 4.2.1.4 Nearby Industrial, Transportation, and Military Facilities 4 20 4.2.1.5 Emergency Planning Zone 4_24 4.2.2 Scope of the Emergency Plan 4-24 4.2.2.1 Emergency Plan Implementing Document 4-25 4.2.2.2 Related TMI Plans, Programs, and Procedure' 4-26 4.2.2.3 Related County and State Plans 4_37 4.3

SUMMARY

OF TMI EMERGENCY PLANNING PROGRAM 4-29 4.3.1 The Emergency Plan 4-29 4.3.2 The Implementing Document 4-31 4.4 EMER0ENCY CONDITIONS 4-32 4.4.1 Licensee Classification System 4-32

4. 4.1.1 Unusual Event 4-33 4.4.1.2 Alert 4-36

4. 4.1. 3 Site Emergency 4 39

4. 4.1.4 General Emergency 4 41 4.4.2 State and County Classification System 4 44

4. 4.2.1 Administrative Events 4 44 4.4.2.2 Emergency Events 4 44 4.4.3 Spectrum of Postulated Accidents 4-45 4.4.3.1 Classification of Hypothetical Accients 4-46

4. 4.3.2 Instrumentation Capability for Detection 4-46

4. 4.3.3 Manpower and Timing Considerations 4-46 4.5 ORGANIZATIONAL CONTROL OF EMERGENCIES 4-48

4. 5.1 Licensee organizations 4-48

4. 5.1.1 Normal Station Organization 4-49 4.5.1.2 Normal Technical Support Organization 4-53

4. 5.1. 3 Gnsite Emergency Organization 4-54

4. 5.1.4 offsite Emergency Support Organization 4-65

4. 5.1. 5 Long-Term Recovery Organization 4-71

4. 5.2 Local Services Support 4-74

4. 5. 2.1 Medical Support Organizations and Personnel 4-74

4. 5. 2. 2 Firefighting Organizations 4-75 4.5.2.3 Law inforcement Agencies 4-75 4.5.2.a Miscellaneous Support 4-75 4-75 4-2 Am. 4 1456 334

TABLE OF CONTENTS Section Topic Page 4.5.3 Coordination 'n'ith Government Agencies 4-75 4.5.3.1 State Agencies 4-76 4.5.3.2 County Agencies 4-79

4. 5.3.3 Federal Agencies 4-82 4.6 EMERGENCY MEASURES 4 85 4.6.1 Act ivation of Emergency Organizations 4-85

4. 6.1.1 Shift Foreman / Control Room Operators 4 86

4. 6.1. 2 Shift Supervisor 4-86

4. 6.1. 3 Duty Section Superintendent 4 37

4. 6.1.4 Emergency Director 4-89

4. 6.1. 5 Dauphin County 4 39

4. 6.1. 6 Pennsylvania Emergency Management 4_gg

4. 6.1. 7 Bureau of Radiation Protection 4, 6. 2 4-90 Assessment Ac t ions 4_gy

4. 6. 2.1 Assessment Actions for Unusual Events 4 97

4. 6.2.2 Assessment Act ions for Alerts 4,9y

4. 6.2. 3 Assessment Actions for Site Emergencies

4. 6.2.4 Assessment Actions for General Emergencies 4-92 4,94

4. 6.3 Corrective Act ions 4,94

4. 6.4 Protect ive Actions 4, 6.4.1 Protective Cover, Evacuation, Personnel Accountability 4-95

4. 6.4. 2 Use of Onsite Protective Equipment and Supplies 4-96 4~9,'

4. 6.4.3 Contamination Control Measures

4. 6.5 Aid to Af fected Personnel 4, 6. 5.1 Emergency Personnel Exposure ,

4. 6.5.2 Decontamination and First Aid - 10 a.

4. 6.5.3 Medical Transportation 4-104

4. 6.5.4 Medical Treatment 4-104

4. 7 EMERGENCY FACILITIES AND EQUIPMENT 4-106

4. 7.1 Licensee Onsite Emergency Centers 4-106

4. 7.1.1 Emergency Control Center ' 4-106

4. 7.1.2 Technical Support Center 4-106

4. 7.1.3 Operations Support Center 4-107 4, 7.2 Licensee Of fsite Emergency Centers 4-108

4. 7.2.1 Offsite Etergency Support Center 4-108 4, 7.2.2 Backup Of fsite Emergency Support Center 4-108

4. 7.3 County, State, and Federal Emergency Centers 4-108

4. 7.3.1 County Emergency Centers 4-10S 4 7.3.2 State Emergency Center 4-110

4. 7.3.3 Federal Emergency Canter 4-110

4. 7.4 Media Center 4-110 4 7.5 Communicat ions Sys tems 4 111 4-3 Am. 4 1456 335

TABLE OF CONTENTS Section Tooic Page 4.7.5.1 Operat ional Line 4-111 4.7.5.2 Radiological Line 4-112 4.7.5.3 NRC Hot-Line 4-113 4.7.5.4 NRC SS4 Line 4-113 4.7.5.5 National Warning System (NAWAS) Line 4-113 4.7.5.6 Pennsylvania Bell Systcm 4-113 4.7.5.7 Microwave System 4-114 4.7.5.8 Radio Communications 4-114 4.7.5.9 Inter-Control Room Hot-Line 4-115 4.7.5.10 Emergency Director's Hot-Line 4-115 4.7.5.11 Alarms 4-115 4.7.5.12 Plant Paging System 4-115 4.7.5.13 Maintenance and Instrumentation Phone System 4-116 4.7.6 Assessment Facilities 4-116 4.7.6.1 Onsite Systems and Equipment 4-116 4.'.6.2 Facilities and Equipment for Offsite Monitoring 4-126 4.7.7 Protective Facilities and Equipment 4-128 4.7.8 First Aid and Medical Facilities 4-131 4.7.9 Damage Control Equipment 4-133 4.8 MAINTAINING EMIRGENCY PREPAREDNESS 4-134 4.8.1 Organizational Preparedners 4-134 4.8.1.1 Training 4-134 4.8.1.2 Drills and Exercises 4-141 4.8.1.0 Emergency Plann;ng Coordinator 4-145 4.8.2 Review and Updating of the Emergency Plan and 4-146 Implementing Document 4.8.3 Maintenance and Inventory of Emergency Equipment 4-148 and SupFlies 4.9 RECOVERY 4-149 4.10 RE FERENCES

                                                                    -151 1456 336 4-4                       Am. 4

4.1 DEFINITIONS The following is a list of terms and their definitions which will be used, as appreeriate, in the Three Mile Island Nuclear Station (Unit 1) Emergency Plan and Implementing Document: 4.1.1 Access Control Point - An access control point serves as the boundary line between the clean and the controllei areas of the plant. The main access control point is located in the Nuclear Services Area, etevation 306.0', in Unit 1. 4.1.2 Accident - An unintentional event which may result in an emergency. 4.1.3 Affected Persons - Persons who, as the result of an accident, have been or may be radiologically exposed or physically injured to a degree requiring special attent ion (e.g. , evacuation, decontamination, first aid or medical services , etc. ) . 4.1.4 Alert - The occurrence of an event or series of events that indicate and allow recognit ion of an actual or potentici substantial degradation the level of safety of the plant. 4.1. 5 Assessment Actions - Those act ions taken during or af ter an accident which are collect ively necessary to make decisions to implement specific emergency actions. 4.1.6 Control Room - The location at Three Mile Island Nuclear Station (TMI) on elevation 355.0' of the Unit 1 Control Building from which the reactor and its auxiliary systems are controlled. 4-8 Am. 4 1456 337

4.1.7 Clean Area - An area where the contamination levels are below those speci-fled below: Loose surface contamination (Beta-Gamma) < 1000 DPM/100cm Loose surface conttmination (Alpha) <100 DPM/100cm Fixed contaminat ion < 0.4 mR/ hour on contact .

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4.1.8 Contaminated Area - An area where the contamination levels are in excess of those specified for a clean area. 4.1.9 Controllec Area - All plant areas where radiation or contamination has a potential for existing in the amounts above the limits set forth for a clean area. 4.1.10 Corrective Actions - Those emergency actions taken to ameliorate or terminate an emergency situation at or near the source of a problem. 4.1.11 Emergency - That situation or condition which may result in damage to property and/or may lead to undue risk to *.he health and safety of the general public and/or site personnel. 4.1.12 Emergency Actions - Those measures or steps taken to ensure that an energency situation is assessed (assessment actions) and that the proper corrective and/or protective actions are taken. 4.1.13 Emergency Action Levels - Predetermined conditions or values, including radiological dose rates; specific contamination levels of airborne, water-borne, or surface-depesited concentrations of radioactive materials; events such as natural disasters or fires; er specific instrument indi-cations which, when exceeded , require the implement at ion of the Emergency Plan-4-9 Am. 4 1456 338

4.1.14 Emergencv Classifications - The characterization of several classes of emergency situations consisting of mutually exclusive groupings including the entire spectrum of possible radiological emergencies. The four classes of emergencies which have been incorporated into this Emergency Plan are (1) Unusual Event, (2) Alert, (3) Site Emergency, and (4) General Emergency. 4.1.15 Em,er,gency Control Center - The designated location from which control and/or coordination of emergency actions is affected. The designated Emergency Control Center for TMI Unit 1 is that area encompassing the Shifts Supervisor's Of fice and the Control Roem both of which are on elevation 355.0' in the Unit 1 Control Building.

4. 1.16 Emergency Core Cooliae System - The Emergency Core Cooling System (ECCS) includes the pertinent pumps, piping, valves, etc. of the Decay Heat Removal System, Core Flooding System, and the makeup portion of the Makeup and Purification System.

4. 1.17 Emergency Director - Designated onsite individual having the responsibility and authority for implementing the TMI Emergency Plan and who will coor-dinate efforts to limit the consequences of the emergency and bring it under control.

4. 1.18 Emergency Operations Center - Designated State, county, and local civil defense headquarters facilities especially designed and equipped for the purpose of exercising ef fect ive coordination and control over disaster operat ions carried out within their jurisdiction.

4.1.19 Emergency Plan - Metropolitan Edison Company plans for coping with emer-gencies at the TMI site, 4-10 Am, 4

                                                                                       ,,n 15

4.1.20 Emergencv Plan Implementing Procedures - Specific procedures in the Implementing Document that include emergency action levels and provide step-by-step emergency act ions . 4.1.21 Emergency Planning Zone - That area, approximately 10 miles in radius around the TMI Nuclear Station, for which emergency planning consideration of the plume exposure pathway has been given in order to assure that prompt and ef fect ive actions can be taken to protect the public and property in the event of an accident. 4.1.22 Emergency Procedures - Specific procedures that provide step-by-step instructions to guide plant operations during potential or real emer-gency situations. 4.1.23 Emergency Support Director - Designated Metropolitan Edison Company / General Public Utili ties employee from the senior management staff who has overall responsibility for accident management and for providing direction and support to the Emergency Director.

4. 1.24 Exclusion Area - As defined in 10 CFR 100.3, that area surrounding the reactor, in which the licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area.

4. 1.25 General Emergency - Accidents which involve actual or imminent substantial core degradation or melting with potential for large releases of radioactive material and/or loss of Reactor Building (containment) integrity, and other accidents that have large radioact ive release potential such as fuel handling and waste gas system accidents.

4-11 Am. 4 1456 340

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4. 1.26 Implementing Document - The document containing an abstract of the TMI Emergency Plan, Emergency Plan Implementing Procedures, and other specific information as required.

4. 1.27 Low Population Zone - As defined in 10 CFR 100.3, the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the eve nt of a serious accident .

4. 1.28 Offsite - Any area outside of the owner property boundary of the Three Mile Island Nuclear Station.

4. 1.29 Onsite - Any area within the owner property boundary of the Three Mile Island Nuclear Station.

4.1.30 Operable - As defined in Technical Specifications (reference 10.13.1), a component or system is operable when it is capable of performing its intended function within the required range. 4.1.31 Owner Controlled Area - That area within the security fence that sur-rounds the immediate site area (i.e. fence that extends from the south vehicle gate north on both shorelines to the north gate). 4.1.32 Personnel Monitoring Equipment - As defined in 10 CFR 20.202, devices designed to be worn or carried by an individual for the purpose of neasuring the dose received (e.g., film badges, pocket dosimeters, thermoluminescent dosimeters, etc.). 4.1.33 Permanent Plant Structure - Those structures that physically house eq uipment , systems, and components tnat are important to planc operat ions 4-12 Am. 4 1456 341

__ _w _ - - (i.e. Reactor, Auxiliary, Fuel Handling, Intermediate, Control, Turbine, and Service Buildings; Process Center; Intake Structure; mechanical and natural draft cooling towers).

4. 1.34 Plume Exposure Pathway - The means by which a radioactive plume can expose the population-at-risk and/or onsite personnel to radiation. The time of potential exposure could range from hours to days. The principal exposure sources from this pathway are: (1) whole body external exposure to gamma radiation from the radioactive plume and from deposited material, and (2) inhalation exposure from the passing radioactive plume.

4.1.35 Population-At-Risk - Those persons for whem protect ive act ions are being or would be taken.

4. 1.36 Population Center Distance - The distance from the reactor to the nearest boundary of a density populated center containing more than about 25,000 residents.

4.1.37 Projected Dose - A calculated or estimated dose which the population-at-risk may potentially receive as a result of a radiological emergency.

4. 1.38 Protected Area - As defined in 10 CFR 73.2, an area encompassed by physical barriers and to which access is controlled. The TMI Nuclear Station protected area includes all areas within the security fence that immediately surrounds the major Station structures (i.e. Reactor, Auxiliary, Turbine, Service, Fuel Handling, and Control Buildings).

4.1.39 Protective Actions - Those actions taken during or af ter an emergency situ-ation that are intended to minimize or eliminate the nazard to the health and safety of the general public and/or site personnel. 4-13 Am. 4 1456 342

4. 1.40 Protective Action Guides - Projected radiological dose or dose commitment values to individuals in the general population that warrant protective action following a release of radioactive material. Prot ect ive act ions would be warranted provided the red c. ion in individual dose expected to be achieved by carrying out the protective action is not offset by exces-sive risks to individual safety in taking the protective action. The protective action guide does not include the dose that has unavoidably occurred prior to the assessment.

4. 1.41 Recoverv Actions - Those actions taken af ter the emergency to restore the plant as nearly as possible to its pre-emergency condition.

4. 1.42 Restricted Area - As defined in 10 CFR 20.3, any area access to which is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. A restricted area shall not include any areas used as residential quarters, although a separate room or rooms in a residential building maybe set apart as a restricted area.

li 1.43 Site Emergencv - Accidents in which actual or likely major failures of plant functions needed for protect ion of the public have occurred. This emergency class includes accident s which have a significant radiat ion release potential.

4. 1.44 State - The Commonwealth of Pennsylvania.

4. 1.45 TBD - An abbreviat ion used to annotate thos e places within this document where specific information or data remains "To-Be-Determined" 4 -1.:, Am. 4 1456 343

4.1.46 Tornado Warning - Meteorological conditions appropriate for a tornado and a tornado sighted in the area. 4.1.47 Unrestricted Area - As defined in 10 CFR 20.3, any area access to which is not controlled by the licensee for the purposes of protectiet of indi-viduals from exposure to radiation and radioactive materials, and any area used for residential quarters. 4 1.48 Unusual Event - The occurrence of an event or events that indicate or allow recognition of a potential degradation of the level of safety of the plant and shall also include contaminated injuries of plant personnel which require of fsite emergency treatment . 4-15 Am. 4 1456 344

4.2 APPLICABILITY AND SCOPE The prime objectives in emergency planning are to (1) provide the means for mitigating the consequences of emergencies (including very low probability events) in order to protect the health and safety of the general public and site personnel and to prevent damage to property and to (2) develop a plan and implementing procedures that will ensure operational readiness and emergency preparedness. This Emergency Plan has been developed for the Three Mile Island Nuclear Station (Unit 1) in accordance with the provision of 10 CFR 50.34, and is consistent with the guidelines given in (1) Regulatory Guide 1.70,

       " Standard Content and Format of Safety Analysis Reports for Nuclear Power Plants", Revision 3; (2) Regulatory Guide 1.101, " Emergency Planning for Nuclear Plants", Revision 1 dated . March 1977; and (3) ". Emergency Planning Review Guideline Number One - Revision One - Emergency Planning Accep-tance Criteria For Licensed Nuclear Power Plants", dated September 7, 1979.

Other guidance and sources of information, which are ident ified in Section 10.0 below, were also used in the development of this Emergency Plan. Although many of the features and requirements detailed in this Plan are, or will be implemented winin a relatively chort period of time following approval, all spects of the Plan will be fully implemented within 60 days prior to the restart of TMI Unit 1. 4.2.1 Applicability of the Emergency Plan 4 2.1.1. General Information and Site Description Three Mile Island Nuclear Station (Unit 1) is operated by Metropolitan Edison Company (Met-Ed), a subsidiary of General Publi: Utilities (GPU). A general area map, provided as 4-16 A** 4 1456 345

Figure 1, shows the relative location of the facility within the State of Pensylvania. The Three Mile Island (TMI) Nuclear Station has a pressurized water type nucl2ar steam supply system supplied by the Babcock & Wilcox Company. It uses chemical shim and control rods for react ivit:- control and generates steam with a small amount of superheat in once-through steam generators. The arrangement of the major TMI Nuclear Station facilities is shown on Figure 2. The TMI Nuclear Station is located in an area of low population density about 12 miles southeast of Harrisburg, Pennsylvania. It is in Londonderry Township of Dauphin County, about 2.5 miles north of the southern tip of Dauphin County, where Dauphin County is coterminal with York and Lancaster Counties. The TMI Nuclear Station site is part of an 814 acre tract consist-ing of Three Mile Island and several adjacent islands which wete purchased by a predecessor company of Metropolitan Edison Company. These lands are part of the York Haven licensed power project (FPC 1888). The TMI Nuclear Station is physically located on Three Mile Island which is one of the largest of a group of several islands in the Susquehanna River upstream from York Haven Dam. The island is relatively flat-land and is wooded on the periphery and on the southern portion. Hills on both sides of the river rise to elevations of over 500 feet. Of the 470 acres which comprise Three Mile island, approximately 200 acres on the northern port ion of the island are set aside for the Station. 4-17 Am. 4 1456 346

The island, which is situated about 900 feet from the east bank and approximately one mile from the west bank of the Susquehanna River, is elongated parallel' to the flow of the r. *r with its longest axis oriented approximately due north and south. The island is approximately 3 miles long and 1700 feet wide and is at an elevation of 300 feet. Its location relative to the other islands in the Susquehanna is shown on Figure 3. Upstream, in the vicinity of Middletown, the southeasterly flowing Susquehanna River makes a sharp change in direction to near due south. After this change in direction, which is just north of Three Mile Island, the river channel widens to approximately 1.5 miles. An access bridge now used by site personnel connects State Highway Route 441 with the north end of the island across Sand Beach Island. A wood access bridge connect s the south end of the island with Route 441 as well. Route 441 is a two lane, black top road which runs north and south parallel to Three Mile Island on the east bank of the Susquehanna River and is more than 2,000 feet from the TMI Unit I reactor at the closest point. The TMI Nuclear Station site is surrounded, except along its southern border, by the Ysrk Haven Pond which is formed by a low dam east and south of the site. This dam does not have locks and there is no commercial water transportation on the river adjacent to the site. On the east bank of the river there is a Penn-Central Railroad one-track line which runs adjacent and 4-18 Am. 4 1456 547

parallel to Route 441. On the west bank of the river, there is a mult i-track Penn-Central line at the river's edge about 1.25 miles west of the site and a black top, two lane road that runs parallel to it . There is a one-track railroad spur across the bridge on tha north end of the island which is used for site-related act ivit ies . 4.2.1.2 Restricted Area. Exclusion Area, and Low Population Zone

1. The restricted area for the Three Mile Island Nuclear Station includes all areas within the owner controlled area. The minimum distance to the restricted area (security fence) boundary is measured from the centerline of the TMI Unit 2 Reactor Building to a point on the western shoreline of the island and is approximately 560 feet.

2. The exclusion area for the Three Mile Island Nuclear Station is a 2,000'f at radius that includes a portion of Three Mile Island, the river surface around it , and a port ion of Shelly Island. The minimum distance of 2,000 feet occurs on the shore of the mainland in a due easterly direct ion . All land areas within the exclusion area are owned by the Metropolitan Edison Company. A map showing the exclusion area boundary is included as Figure 4

3. The low population zone for the Three Mile Island Nuclear Station has a minimum distance of 2 miles to its outer boundary. The area of the low population zone is also shown on Figure 4 4-19 Am. 4 1456 348

'.2.1.3 Pooulation and Population Distribution As discussed in the previous subsect ion, a low population zone with a distance tc it s outer boundary from the TMI Nuclear Station has been specified to be 2 miles. The nearest population center is Harrisburg, Pennsylvania (1970 population of 68,000) which is located approximately 12 miles northwest of TMI. This distance satisfies the requirements of 10 CFR s0 with respect to population center distance. The present and projected population distribution as a function of distance and direction from the TMI Nuclear Station based on 1970 census is shown on Figure 5. The counties and relatively large population centers are further detailed on Figure 6. The population of surrounding residential areas, the enrollment in tha various schools, and the number of patients in the hospitals in the area around the TMI Nuclear Statica site are tabulated in Tables 1, 2, and 3 respect ively. There are no schools and only one recreational area (Falmouth Fish Comm;ssion) presently within the 2 mile low population zone. There is some seasonal shift in population within a 5 mile radius of Three Mile Isand since there are over 100 summer cabins on the islands within the area. Additional transients part ic ipate in boating activities in the vicinity of the TMI si.e. 4.2.1.4 Nearbv Industrial, Transoortation, and Military Facilities

1. The Three Mile Island site is currently surrounded by farm lands within a 10 mile radius. Lands are used for 4-20 Am. 4 1456 349

dairy cattle, tobacco, poult ry, vegetables , fruit, corn, wheat, and other products. A summary of land use for Daushin County, ... which the TMI site is located, and for the two other nearest counties (York and Lancaster) is provided on Table 4. The Susquehanna River is used for sport fishing and boating and is not used for commercial fishing.

2. Manufacturing industries in the region produce clothing, wood products, shoes, electrical wiring devices, steel products, packed meat and other food. These act ivit ies ,

within a 10-mile radius of the site, are confined chiefly to the cities of Harrisburg, New Cumberland, Steelton, and Middletown. A listing of typical industries within 10 miles of the TMI site and the.r respect ive employment is summarized on Table 5. Located within a 5 mile radius of the site are approximately 23 industrial firms that employ about 2,400 persons. There are gas and oil transmission lines located at a mininum distance of approximately 2 miles from TMI. The only power station in the immediate upstream vicinity of TMI is the Crawford Station (retired) which is owned by Metropolitan Edison Company. Crawford Station is located about 2.5 miles upstream from the site on the east bank of the Susquehanna River. I= mediately downstream from the site is the York Hiien hydroelectric project, consisting of the main dam which averages approximately 10 feet in 4-21 Am. 4 1456 350

neight and extends about 4,970 feet across the main river channel to Three Mile Island; a secondary dam which is about 8 feet high and extends 950 feet across the east channel of the river; a pool extending approximately 3.5 miles upstream from the dam and contains about 10,000 acre-feet of volume; and a head race wall that is about 20 feet in height extending from the west end of the main dam appror.imately 3000 feet to the plant. The York Haven Station is operated on a "run-of-the-river" basis, and its power output is dependent primarily upon the water available. However, the reservoir is used for peaking operat ion during periods of low ritar flow. Brunner Islanc Station, a large steam-electric generating plant owned by the Pennsylvania Power & Light Company is located on the Susquehanna River

,e     imately one mile downstream from the York Haven pro s ect. This station uses water from the river on a "run-through" basis for cooling water.      Three other hydro-electric generating stat ions are also located downstream from TMI, with each project having a dam and reservoir on the Susquehanna River. The three stations are Safe Harbor, Holtwood, and Conewingo Hydroelectric project s , located approximately 25, 31, and 47 miles south of Three Mile Island, respectively. There is also a coal fired, steam-electric plant at Holtwood, and the Muddy Run Pumped Storage Project is :4ssociped with Conewingo Station. The Peach Bottom Nuclear Generating Station is located along the west bank of the Sascuehanna River, about 41 miles 4-22
                                                              ;5' 1456

downstream of Three Mile Island, just north of the Maryland-Pennsylvania border and is the only other nuclear plant within a 50-mile radius of Three Mile Island.

3. There are two airports within 10 miles of the TMI site.

Harrisburg International Airport (formerly Olmsted Air Force Base) is located on the north bank of the Susquehanna River approximately 2.5 miles northwest of the site. The Capital City Airport is located appreximately 8 miles west-northwest of the TMI site. It is important to mention that vital areas of the TMI Nuclear Station were designed to withstand a hypothetical aircraf t incident. Penn-Central Railroad Lines are located on both sides of the Susquehanna River, the closest being the east bank, approximately 2,000 feet from the Three Mile Island Unit 2 Reactor Building. Rout ine traf fic in liquified petroleum gas was identified on the railroad line which passes along the east shore of the river. Analyses indicate that any missiles generated by this traf fic would be less damaging than the postulated aircraf t strike agcinst which the plar.t is protected and that flammable gases would dissipate before reaching the TMI Nuclear Station.

4. The closest military installation to the site is the Air National Guard facility at Harrisburg International Airport.

There are no military firing ranges or missile facilities within a 10 mile radius of TMI. Other military facilities, 4-23 Am. 4 i456 35n2 e

however, are Army and Navy depots located respectively at New Cumberland and Mechanicsburg, Pennsylvania. A map showing the major industrial and military facilities within a 5 mile radius of TMI has been included as Figure 7.

4. 2.1. 5 Emergency Planning Zone Metropolitan Edison Company has, in defining the Emergency Planning Zone (EFZ) for Three Mile Island Nuclear Station, taken into consideration the information and data presented above as well as other important factors such as organizational capabilites, availability of emergency facilit ies and equipment , and the methods for implement ing the TMI Emergency Plan. As such, an EPZ having a minimum radial distance of 10 miles from the TMI site has been defined. A map of the area around TMI that shows the boundary of the EPZ is provided as Figure 8.

4.2.2 Scope of the Emergency Plan Total emergency preparedness requires plans for the emergency response of both systems and people. The engineering design of the TMI Nuclear Station ensures that the consequences of major malfunctions will be mitigated by the engineered safety systems. The purpose of the TMI Emergency Plan is to provide the basis for human emergency response in the way that design does for system response. To further ensure that the response to emergencies is initiated in a timely manner and ef fect ively controlled, it is of prime importance that emergency plans be coordinated with other plans, programs, and procedures such as those discussed below. 4-24 A** 4 1456 353

4.2.2.1 Emer?ency Plan Imolementine Document The TMI Emergency Plan has a separate Implementing Document which is distributed to those individuals, agencies, organiza-tions, and facilities where immediate availability of such information would be required in an emergency. The Implementing Document table of content s is attached as Appendix A to allow for review of its organization and the list of Emergency Plan Implementing Procedures. As can be seen by reviewing Appendix A, the Implement ing Document is organized to provide:

1. An abstract of the TMI Emergency Plan which will include such specific items as definitions, discussions of emergency organizations and responsibilities, a description of emergency planning related systems and facilities, lists of emergency equipment , and other information such as important charts and maps.

2. Detailed Emergency Plan Implementing Procedures will, as necessary and appropriate, be used to assess conditions, classify the emergency, provide directions for making required notifications and for requesting assistance, and provide the step-by-step instructions for init iating protect ive and correct ive act ions .

The Emergency Plan Implementing Procedures, have a direct relationship to the TMI Emergency Plan and are coordinated with several other TMI Nuclear Station plans, programs, and 4-25 ^* ' 1456, 354

procedures including those discussed in subsect ion 2.2.2 below. 4.2.2.2 Related TMI Plans, Programs, and Procedures

1. The TMI Security Plan and proc?dures and the TMI Emergency '

Plan and its Implementing Procedures have been coordinated to ensure that appropriate emergency actions can be taken. For example, the Security Plan and procedures will have provisions for emergency response personnel and vehicle access when required by the Emergency Plan Implementing Procedures.

2. The TMI Radiation Protect ion Manual defines administrative controls and procedures such as radiological control limits and precaut ions , use of personnel monitoring devices , use of protect ive clothing and equipment, personnel decontamin-ation, etc. In addition, Health Physics Procedures provide instruct ions on performing surveys , analyzing samples ,

operat ing health phys ics/radiat ion protect ion equipt ent , etc. The pertinent information and details rea: ced in these documents have either been incorporated into the TMI Emergency Plan and/or Implementing Procedures or have been appropriately referenced.

3. A comprehensive set of Emergency Procedures that are used to control plant operations during abnormal and accident conditions have been prepared. Since there is a direct relationship between emergency operat ions and emergency 4-26 Am. 4 1456 35e3

planning, it is of prime importance that F.mergency Pro-cedures and Emergency Plan Implementing Procedures be closely coordinated and complimentary. Because of this, specific Emergency Procedures will, as appropriate, direct the on-shift operations personnel ta the applicable Emergency Plan Implementing Procedure (s). Conversely, Emergency Plan Implementing Procedures will ensure that applicable Emergency Procedures are ut ilized when appropriate. 4 Metropolitan Edison Company has developed an Emergency Public Information Plan for the TMI Nuclear Station. The purpose of this Communications Plan is to describe the methods by which Met-Ed will release information to the

                  =edia and the public in addition to the internal dis-seminat ion of information in the event of emergencies at TMI. This Communications Plan has been included in the development of the TMI Emergency Plan and is attached as Appendix B.

4.2.2.3 Related County and State Plans The development of the State plans and the TMI Emergency Plan were closely coordinated. In addition, specific State require-ment s for report ing of emergencies , providing info r=at ion and data, recommending protect ive act ions , etc. , have been integrated directly into the Emergency Plan Implementing Procedures. In considering The Emergency Planning Zone, there are f.ve county plans (i.e. .auphin, York, Lancaster, Cumberland. and Lebanon) that have been factored into the development of the TM! Emergency 4-27 Am. 4

                                                                          ) k )b

Plan. It is impo rt ant to point out that not only is the TMI Emergency Plan coordinated with the State and County plans, but the State and County plans are coordinated as well. The details of the State and County plans are provided in subsection 5.3.1 below. ~ ' ' - ~ - ..- - . . 4-28 Am. 4 1456 357

4.3

SUMMARY

OF TMI EMERCENCY PLANNING PROGRAM The TMI Emergency Planning Program, as defined by the Metropolitan Edison Company, consists of two separate but totally coordinated documents. The first document , this Emergency Plan, provides the meaue for performing advance planning and for defining specific requirements and commitments that will be implemented by other documents and procedures (eg. Admini-strative Procedures, Surveillance Procedures, Emergency Plan Implementing Procedures, etc.). The second document, the Emergency Plan Implementing Document, provides the detailed information and procedure = that will be required to implement the TMI Emergency Plan in the event of an emergency at the TMI Nuclear Station. Thest two documents are briefly described below. 4.3.1 The Emergency Plan The TMI Emergency Plan assures that all emergency situations, including those which involve radiation or radioactive ::..;terial are handled logically and efficiently. It covers the entire spectrum of emergencies from minor, localized emergencies to major emergencies involving action by of fsite emergency response agencies e-d organizat ions . The TMI Emergency Plan includes a scheme for classifying emergencies that meets the current guidance (reference 10.20) provided by the Nuclear Regulatory Commission (NRC). This classification system is described in detail in Section 4.0 below. Furthermore, this Plan incorporates specific response criteria (emergency action levels) which will be used in the assessment of emer-gency situations. Thus, the TMI Emergency Plan provides the overall advance planning required for the development of methods of iroplementation which will be included in the Implement ing Document . 4-29 Am. 4 1456 358

4.3.1.1 In. summary, the TMi Emergency Plan provides:

1. A means for classifying emergency conditions in a manner compatible with a system utilized by State and County emergency response agencies and organizations.

2. A means of reclassifying such emergency conditions should

• the severity increase or decrease.

3. Deteils of normal and emergency operat ing organizations.

4. General guidelines as well as specific details as to which State, County, and federal authorities and agencies and other out side organizations are available for assistance.

5. Information pertaining te the emergency facilities and equipment available both onsite and offsite.

6. Guidance for the preparation of detailed Emergency Plan Implement ing Procedures.

7. Requirement s, such as training, drills, reviews, and audits, which will result i t. a high degree of emergency preparedness and operat ional readiness.

8. Figures and tables which display detailed information and data such as organization charts, maps, population distri-but ions , etc.

9. An appendix det ailing specific plans and agreement s pertain-ing to participating ef fsite organizations and agencies.

4-30 Am. 4 A56 559

                                                  .   --         . ~ - , _            ~--

4.3.2 The Imolementing Document The purpose of the Emergency Plan Implementing Document, a descript ion of which is attached as Appendix A, is to provide a " single source" of pert inent and significant information and data and the procedures that would be required by or useful to various emergency response agencies and organizations in the event of an emergency at TMI Nuclear Stat ion. The I=plementing Document, therefore, consolidates and integrates specific material detailed in such documents as the TMI Emergency Plan, the State Plans, and the various County Plans. 4.3.2.1 The Implement ing Document is organized to provide:

1. An abstract of the TMI Emergency Plan, including specific items sucn as definitions, emergency organizations, responsi-bilities, facilities, equipment, and means of classifying emergencies.

2. Emergency Plan Implementing Procedures that define specific emergency action levels which rtquire implementation of the procedure (s) and the detailed emergency actions (i.e.

step-by-step instructions) of the procedure (s). 4-31 Am. 4 1456 % 0

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4.4 EMERGENCY CONDITIONS 4.4.1 Licensee Classification System This Emergency Plan characterizes four classes of emergency situations which consist of mutually exclusive groupings covering the entire spectrum of possible emergency situations. Each class is associated with a particular set of immediate actions to be taken to perform (1) accident classification, (2) notification of offsite agencies and support groups, and (3) mobilization of the applicable port ion of the emergency organiza-tions to cope with the situation and continue accident assessment functions. These actions are described in deta 3 in Section 6.0 of this Plan. The various classes of accidents represent a hierarchy of accidents based on potential or actual hazards presented to the general public. Accidents may b = classified in a lower category at first and then escalate to another, higher class if the situation deteriorates. Each of the four emergency classes are characterized by emergen . o act ion levels . These levels consist of specific sets of plant p. ;ers (i.e., instrument indications, system status, etc.) that will be used to initiate (1) emergency class designation, (2) not ificat ion, and (3) emergency organization's mobilization. It must be noted that these emergency action levels are used specifically to obtain early readiness status on the part of emergency response persons and organizations. These levels have not been selected so as to infer any immediate need to implement protective actions but rather to insure a reasonable amount of time is available to confirm in plant readings by implement ing assessment measures onsite and of fsite. Once declaration of an emergency class requiring possible protective action occurs, dose assessments will be 4-32 ^"* ' 1457 001

made by either measurement or project ion methods. The dose assessment values, along with other plant status assessments will be reported to offsite agency officials as inputs for their decision on whether or not protective actions for the public are to be implemented. The relationship of these dose assessment values to the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) and the possi-bility of approaching or exceeding the PAGs will specifically be reported. The philosophy taken for classification will always be to immediately declare the highest class for which an emergency act ion level has been exceeded. For example, a Site Emergency would be declared directly if a Site Emergency actien level is exceeded, without having previously been in the lower, Alert class. Each of the four emergency classes, and the related emergency action levels requiring declarat ion of each class, are described in the following subsections. 4.4.1.1 Unusual Event The least severe of the four emergency classes defined by this Plan is criled an Unusual Event. For the purposes of this Plan, an Unusual Event shall be defined as the occurrence of an event or event s that indicate or allow *ecognition of a potential degradation of the level of safety of the plant. For convenience, this class shall also include contaminated injuries of plant personnel which require offsite emergency treatment. The incident shall be classified as an Unusual Event only if the event is a minor one and no releases of radioactive material 4-33 Am. 4 1457 002

requiring of fsite response or monitoring are expected. Events in this class are selected based upon a potential to degenerate to a more severe situation rather than an actual public hazard. The emergency action levels that shall require an Unusual Event declaration include (but are not necessarily limited to) the following:

1. An automatic reactor trip coincident with either:

a. ' oss of Reactor Coolant Flow

b. Loss of main and auxiliary feedwater

2. Any reactor trip followed by an unanticipated automatic ECCS actuation.

3. Reactor Building pressure > 2.0 psig.

4. Reactor coolant total activity > 50 uCi/ml.

5. Any unidentified Reactor Coolant System leakage > 1 gpm or total Reactor Coolant System leakage > 10 gpa that results in a Technical Specificatien required shutdown.

6. Sustained loss of all of fsite power which results in a reacto: trip.

7. Both Diesel Generators inoperable resulting in a Technical Specification required shutdown.

8. Projected river stage > 302 ft. at the River Water Intake Structure.

9. Any earthquake of a magnitude > .Olg as indicated by the

     " Threshold Seissic Condition" annunciator.

4-34 Am. 4 1457 003

10. Trans portat ion of a contaminated , injured individual from onsite to an offsite hospital.

11. Actual or projected hurricane force winds (> 75 mph sus-tained).

12. Onsite aircraf t crash outside the protected area fence and not impacting on plant structures.

13. Any near or onsite toxic or flammable gas or liquid release which affects the habitability required for normal plant operations.

14 Valid alert alarm on an effluent radiation monitor gas channel. (Monitors RM-A8, RM-A9 and RM-AS)

15. Any fire in a permanent plant structure which cannot be controlled by the Fire "rigade within 10 minutes of discovery or any fire outside plant stuctures requiring of fsite firefighting assistance.

16. Any valid Reactor Building evacuation alarm.

The intent of the values noted above is to provide absolute values which, if exceeded, will initiate the Unusual Event emergency clas s. In addition to the requirements for declaration of this emergency class that are imposed by the emergency action levels described above, the Unusual Event class can be declared by an action statement in a specific Emergency Procedure er Alarm Response Procedure. Steps in these procedures state that an Unusual Event has occurred or is occurring and require that an Unusual Event i457 004 4-35 Am. 4

class of emergency be declared. All Emer;ency Plan related act ions (not ificat ion, etc.) will be carried out in parallel with the remainder of the Emergency Prccedure. Lastly, the imergency Director shall declare an Unusual Even; any time that , in his judgement , the plant status warrants such a declaration. Training shall stress the need to analyze all minor events in light of their potential for further degradation of the level of safety of the plant and not hesitate to declare this particular emergency clans. 4.4.1.2 Alert The next level of emergency class designated in this Plan is called an Alert. An Alert is the occurrence of an event or series of events that indicate and allow recognit ion of an actual or potential substantial degradation of the level of safety of the plant. As in the case of an Unusual Event, the Alert class includes emergency situations that are expected to be minor in nature but where it has been deemed prudent to notify more of the offsite emergency participants and mobilize a larger portion of emergency organization. In addition, because of the nature of the Alert class, (releases of radio-active material possible) broader assessment actions will be started. All of the act ions to be taken for each emergency class -: described in detail in Section 6.0 of this Plan. Events that will initiate an Alert shall be those with the potential of limited releases of radioactive material to the environment. As before, a situation will only be classified 4-36 Am. 4 1457 005

at the Alert level if none of the emergency action levels for a tigher class have been exceeded or are expected to be exceeded in the near term. The emergency action levels that shall require an Alert to be declared include (but are not necessarily limited to) the following:

1. Reactor Coolant System pressure and temperature reach saturation conditions.

2 Reactor Coolant System hot leg temperature > 620 F.

3. Reactor thermal power and reactor power imbalance in excess of the safety limits defined by Figure 2.1-2, " Core Protection Safety Limits", in reference 10.13.1.

4. A measured Reactor Coolant system pressure in excess of 2500 psig.

5. Failure of the Reactor Coolant System power operated relief valve to shut (af ter lifting to relieve pressure) .

6. Reactor Building pressure > 4.0 psig.

7. Reactor coolant total activity > 130 uCi/ml.

8. Primary to secondary system leakage in excess of I gpm.

9. More than one contol rod known to be untrippable.

10. Loss of all offsite power coincident with loss of both Diesel Generators,

11. Secondary system act ivity (I-131 equivalent) > 1.0 uCi/ml.

12. Actual river stage > 302 f t. at the River Wate.r Intake Structure.

Any earthquake of a magnitude > OBE levels as indicated by an alarm on the " Operating 3 asis Earthquake" annunciator. 1t. m nade warning. 1457 006 4-37 Am, 4

15. An aircraft crash within the protected area or onto any permanent plant structure.

16. A fire in any permanet.: plant structure which requires offsite firefighting capability.

17. A valid count rate on any gaseous plant effluent monitor that would result in a projected dose rate at the exclusion area boundary of'> 10 mR/hr (gamma) usiag adverse meteorologv.

18. A valid count rate on any plant iodine effluent monitor that would result in a projected child thyroid dose at the exclusion area boundary of 2,50 mR in one hour using adverse meteorology.

19. A valid dose rate on the Reactor Building high range monitor that would result in a projected child thyroid dose at the exclusion area boundary of 2,50 mR in one hour or a whole body dose rate 2,10 mR/hr (gamma) using adverse meteorolorv and building design leakrate.

20. Unanticipated high alarm on any two area and/or process radiation monitors at the same time.

21. A high alarm on the Station liquid effluer.t monitor (RML-7).

Again, the values specified are absolute act ion levels requiring declaration of the Alert level. This class of emergency can alsc be declared by arrival at an act ion statement in a specific Emergency Procedure. Steps in these procedures state that an Alert has occurred or is occurring and require that an Alert class of emergency be declared. All Emergency Plan related act ions (not i ficat ion , etc.) will be 1457 007 4-38 Am. 4

carried out in parallel with the remainder of the Emergency Procedure. As in all cases, the Emergency Director shall declare an Alert any time he judges the status of the plant to warrant it. He shall specifically consider escalation from the Unusual Event to the Alert class if, in his judgement, the situation is not likely to be resolved rapidly or is likely to deteriorate. 4.4.1.3 Site Emergency The next highest level of emergency class designated is the Site Emergency. The Site Emergency class includes accidents in which acti al or likely major f ailures of plant functions needed for protection af the public have occurred. Although immediate protactive actions are not automatically required, declaration of a Site Emergency will set into motion all personnel onsite and offsite that would be required to r rferm actions up to and including the evacuation of near-site areas. All monitoring teams required to =ske e,atinuing assessments for providing of ficials with information to decide on protective actions will

b. dispatched. The Site Emergency class includes accidents which have a significant radiat ion release potential. Details of all of the emergency measures that will be taken upon declar-ation of a Site Emergency are presented in Sectior 6.0 of this Plan. The emergency action levels that shall require a Site Emergency to be declared shall include (but are not necessarily limited to) the following:

1457 008 4-39 A '

1. Reactor Building pressure > 30 psig.

2. Reactor coolant activity > 300 uCi/ml.

3. Primary to secondary leakage > 50 gpm.

4. Loss of all offsite power and loss of both Diesel Gener-ators coincident with total loss of vital AC and DC power.

5. River stage > 307 ft. at the River Water Intake Structure.

6. Any earthquake of magnitude > SSE levels.

7. Failure of actuated Emergency Core Cooling System components to start and run following an automatic system initiation such that the number of compenents available is below the minimum assumed in the accident analysis.

8. A valid count rate on any gaseous plant effluent monitor that would result in a projected dose rate at the exclusion area boundary of > 50 mR/hr (gamma) using adverse meteorology.

9. A valid count rate on any plant iodine effluent monitor that would result in a projected child thyroid dose at the exclusion area boundary of > 250 mR in one hour using adverse meteorology.

10. A valid dose rate on the Reactor Building high range monitor that would result in a projected child thyroid dose at the exclusion area boundary of > 250 mR in one hour or a dose rate > 50 mR/hr (gamma) using adverse meteorology and building design leakrate.

11. Incore temperature > 700 F as measured by any two incore thermocouple readings following a reactor trip.

12. Offsite radiological monitoring reports of > 50 mR/hr (gamma) at any location.

1457 009 4-40 Ac. 4

This emergency class, as is the case with all classes in this Plan, can also arise from an action statement in a specific Emergency Procedure. As an example, one of the steps in the

       " Loss of Reactor Coolant / Reactor Coolant Pressure" Emergency Procedure will trigger declaration of a Site Emergency.

In addition, the Emergency Director shall declare a Site Emer-gency any time that in his judgement plant conditions exist that warrant the activation of emergency centers and precautionary public notification. It should be noted that, unlike the two previously described classes of emergency, the Site Emergency class is very likely to involve some radiat ion expostre to the near-site public. Also, many of the accidents included 'in the class have the potential for degradation to the General Emergency class. Although the emergency action levels for this class have been selected at values well below the protect ive act ion guides, of fsite moni-toring team reports and continuing assessment will lead to any final decision on protect ive act ions to be taken. A detailed discussion of the implementation of assessment ct ions and possible i.rotect ive act ions for the Site Emergency class are provided in Sect ion 6.0. 4.4.1.4 General Emergency The highest, most severe class of emergency defined by this Emergency Plan is called General Emergency. Tne General Emergency class will contain accidents which in"olve actual or imminent 4-41 Am. 4 1457 010

substantial core degradation or melting with potential for large releases of radioactive material and/or loss of Reactor Building (containment) incegrity, and other accidents that have large radioactive release potential suen as fuel handling and waste gas system accideats. In keeping with the philosophy adopted througheut this 21an, the emergency action levels ~are set at values below the EPA protective action guides so that they may be used to (1) declare the emergency, (2) notify the appropriate authorities and support groups and (3) mobilize the applicable portions of the emergency organizations. However, this clasa of emergency is somewhat different in that some protective actions may be recommended upon declaration of the General Emergency, since, for this emergency class, the lower limits of the protective action guides are likely to be exceeded. The emergency act ion levels have been selected so that time should be available to make some confirmatory measurements in the field prior to implementation of any of the more extensive (i.e., evacuation) protective actions. Some of the General Emergency acticn levels require a dose projection calculation using actual meteorology. This differs from the adverse meteorology assump-tions used in the Site Emergency action levels in order to remove this built-in conservatism and to preclude declaring a General Emergency when actual conditions do not warrant that high of classification. In those cases where radiation monitors are of fscale, the calculations will use contingen:y dose release 4-42 Am. 4 1457 011

factors. A detailed description of the protective actions to be taken for each class of emergency will be presented .n Sect ion 6.0. The- emergency action levels that shall require declaration of a General Emergency include (but are not necessarily limited to) the following:

1. A projected dose rate at the exclusion stea boundary from all sources of > 100 mR/hr (gamma) using actual meteorology and Reactor Building design leakrate.

2. A projected child thyroid dose from all sources at the exclusion area boundary of > 500 mR in one hour using actual m.steorology and Reactor Building design leakrate.

3. Offsite monitoring reports of > 100 mR/hr (gamma) at any location.

4. Reactor Building high range monitor exceeds the level for a Site Emergency with either (1) a measured Reactor Building pressure > 55 psig or (2) a measured hydrogen concentration in the Reactor Euilding > 3 by volume.

Also, the Emergency Director shall declare a General Emergency at any eime that in his judgement plant conditions exist that already, or in the near term warrant the taking of sone pro-tect ive act ions for the population-at-risk. Although the emergency action levels are set below values related to the EPA protective action gudies to allow time for further assess-ment, it is likely that some protective actions will be required for a complete discussion of these actions refer to Section 6.0. 4-43 Am. 4 1457 012

4.4.2 State and County Classification System The Department of Environmental Resources, Bureau of Radiat ion Protect ion, Plan for Nuclear Power Generating Station Incidents (included with Appendix D) characterizes two classes of emergency situations; Administrative Events and Emergency Events.

4. 4.2.1 Administrative Events As stated in the Bureau of Radiaticn Protection (BRP) Plan, Administrative Events "are those which have little or no radiation consequences to the public but may be of significant public interest." By the BRP's definition, Administrative Events may include, but are not necessarily be limited to, such situations as (1) personnel emergencies, (2) any event which leads to exceeding a radiological report level in measured environmental parameters specified in reference 10.12.2, and (3) any unplanned event at the TMI Nuclear Station wnich prompts Met-Ed to issue a press release.

For the purpose of relating the State's emergency classification system to the classification system defined in Section 4.1 above, a direct correlation between the BRP's Administrative Event and Met-Ed's Unusual Event characterizations shall be made. 4.4.2.2 Emergenes Events As stated in the BRP's Plan, an Emergency Event is any condition or event which has the potential to discharge significant quantitiea of radioactive material to the public domain. It also obviously includes actual discharge." 4-44 Am. 4 1457 013

For the purpose of relating the State's emergency classification system to the classification system defined in Section 4.1 ab ove , a direct correlation between the BRP's Emergency Event and Met-Ed 's Alert , Site Emergency, and General Emergency charact erizat ion. shall be made. Since both the State's Emergency Event classification and Met-Ed's Site aad General Emergency classification's inicude events which have significant potential for radioactive releases, it is imperative that specific guidance for initiating pro-tective actions be available to the " decision-making" personnel in emergency response organizations and agencies. The State has, for planning purposes, auopted the Environmental Protection Agency's (EPA) protective act ion guides (PAG's) that are speci-fled in reference 10.8. It is impo rt ant to mention that the projected values for dose and dose commitment given as emergency action levels for even the highest class of mergency (i.e. General Emergency) are considerably lover than the EPA PAG's discussed above. Therefore the declaration of a Graeral Emergency, although an extremely significant event in it s own right, should not be construed to mean that the EPA PAG's have or even will be exceeded. 4.4.3 Spectrum of Postulated Accidents This sect ion of the TMI Emergency Plan shows that each of the discrete accidents that have been hypothesized for the plant is encompassed 4 thin the preceding emergency characterization classes and provides a summary analysis of their implications for emergency planning. 4-45 Am. 4 1457 014

4.4.3.1 Classification of Hypothetical Accidents All of the events hypotbesized in Chapter 14 of the TMI Nuclear Station Final Safety Analysis Report (FSAR) fall into one of the four emergency classes outlined above, with approximately half falling into the Alert, Site, and General Emergency categet ies. Table 6 lists eac;. of these events and the emerger.cy class that each would be likely to fall into. A complete discussion of any of these hypothetical events may be found in Chapter 14 of the FSAR. It must be noted that in completing this table the most conservative accidents described in Chapter 14 have been assumed. Therefore, occurrence of some of these accidents (for example with no failed fuel) may not result in as high a class as noted in Table 6. Also, equipment assumed to work in the

                                                                                              ~

Chapter 14 analysis were assumed to successfully operate for the evaluation. Failures of required equipment in any of the accident scenarios may result in higher classes of emergencies. 4.4.3.2 Instrumentation capability for Detection The plant instrumentation that will be used to promptly detect accidents at the plant is discussed in detail in the TMI Nuclear Station FSAR. Table 7 lists each hypothetical accident, and the impo rt ant instrumentation that would be expected to detect each of these accidents. Only major, installed equipment is listed. 4.4.3.3 Manoower and Timing Considerations The manpower response and timing considerations for the four emergency classifications are depicted in Table S. Th i s t ab le A** ' 4-46 1457 015

includes (1) the number of personnel onsite continuously, (2) the number of personnel to be called to report onsite, and (3) those that are to be called to report to the Off-site Emergency Support Center. Included in Table 8, along with the above information, is the estimated maximow time for the identified personnel to report to their assigned locations. 4-47 Am 4 1457 016

4.5 ORGANIZATIONAL CONTROL OF EMERGENCIES In preparing this Jection of the TMI Emergency Plan, a slight deviation from the format used in reference 10.4 was deemed necessary in order to present the Met-Ed/GPU organizations (normal and emergency) and the various support organizations in a logical manner. It is impottane-to ment ion that each of the content-related requirements of ceference 10.4 for this Section is addressed in the following subsections. To facilitate the review of thi.s section against reference 10.4 criteria the following cross index should be used: o Subsection 5.1.1 below was prepared in accordance with subsection 5.1 of reference 10.4. o Subsection 5.1.2 below was prepared in accordance with subsection 5.2 of reference 10.4 o Sub s ec t io.:s 5.1.3, 5.1.4. and 5.1.5 below were prepared in accor-dance with subsect ion 5.3.1 of reference 10.4. o Subsection 5.2 was prepared in accordance with subsection 5.3.2 of reference 10.4. o Subsect ion 5.3 was prepared in accordance with subsection 5.4 of reference 10.4. 4.5.1 Licensee Organizations The TMI Generation Group is the organization which operates and provides technical support for the TMI Nuclear Station. This organization is staffed by Met-Ed/GPU personnel. The following five subsections provide a detailed description of the TMI Generation Group Station and Technical Support Organizations during normal operations and, in addition, a detailed descriptier of the onsite and of fsite emergency organizations Am. 4 4-48 1457 017

that can be activated from the normal organizational artangements. A 4 descript ion of a basic organization for long-term recovery operations is also provided. 4.5.1.1 Normal Station Organization

1. A block diagram of the TMI Nuclear Station organization is provided as Figure 9. The diagram illustrates the levels of responsibility within the Station organization.

The personnel staf fing the nortal Station organization are usually otsite from about 8 AM to 5 PM during the normal work week (i.e., weekends and holidays excluded),

a. In the normal Station organization, the Unit 1 Super-intendent shall be responsible for the overall safety of plant operations. He shall report directly tc the Vice President of Nuclear Operations and is responsible to him for the operation, and preventive maintenance of TMI Unit 1.

b. The Director of Technical Support shall be responsible for providing engineering and analysis support ser-vices for the Station. In addit ica, he i. charged with the responsibility for providing a technical advisor to each operations shift. The Director of Technical Support also reports to the Vice President of Nuclear Operations.

c. The Manager of Training reports to the Vice President of Nuclear Operations and is responsible for the 4-49 Am. 4 1457 018

conduct of Station training programs such as general employee training, basic health physics training, licensed operator training and requalification, technician t:aining, and training related to emergency planning.

d. The Manager of Support Services and Logistics has responsibility for the administration of the Station.

In addition, he also oversees the Site Security Force and provides for health physics and chemistry support s e rv ic e s . He also reports to the Vice President of Nuclear Operat.ons.

2. The Operations Department provides operators onsite on a rotating shift basis to ensure the safe and proper operation of the plant 24 hours per day, 7 days per week. In addition, personnel from other departments within the station organi-zation are also assigned to shifts to provide additional capabilities. Requirement s for minimum shif t crews are specified in Section 6.2 of the Technical Specifications (reference 10.13.1), however, the typical TMI Unit I shift organization is shown on Figure 10. I* is important to ment ion that the shif t organization can be augmented, in the event of an emergency, with of f-duty personnel within 60 minutes.

a. The Shif t Supervisor, who is on duty at all times, is in the immediate onsite position of authority and 4-50 Am. 4 1457 019

responsibility for the safe and proper operation of the plant. The Shift Supervisor will be responsible for the initial evaluation of any abnormal or emer-gency situation and for directing the appropriate response. If it is determined that an emergency exists, those responsibilities assigned to the Emergency Director will be assumed by "ne Shift Superviser. Under such circumstances, the Shift tupervisor will initiate appropriate actions, imple-ment proper procedures, notify appropriate offsite emergency response organizations and agencies (i.e. Dauphin County, PEMA, NRC) and the Duty Sect ion Superintendent, and retain such responsibilities until relieved a. the Emergency Director. The Shif t Super-visor shall, during normal and emergency operatic.'s, maintain control over plant operations as the senior licensed operator unless he is properly relieved by another member of the TMI Nuclect Station staf f who holds a valid Senior Operater license. In addition, the Shift Supervisor shall maintain control over the conduct of operations and personnel in the Control Room.

b. The Shift Foreman functions as an assistant to the Sh i f t Supervisor. In the temporary absence of the Shift Supervisor, the Shift Foreman will assume his responsibilities.

4-51 Am. 5 1457 020

c. The Control Room Operators are responsible for the manipulation of controls as necessary to perform plant operations as directed by the Shift Foreman or Shift Su pe rvi s or. They are responsible to the Shift Foreman.

d. The Auxiliary Operators are responsible for performing component and/cr system operations outside of the control room. Normally they take direction from the Control Room Operators, however, they too are respon-sible to the Shift Foreman or Shift Supervisor.

e. In addition to the Operations personnel assigned to each stift, a Shift Technical Engineer from the Technical Support Department will be assigned to each shift. He will be responsible for serving in an advisory capacity on plant safety to the Shif t Su pe rvi s o r . He shall have no duties or responsi-bilities for manipulation of controls or for command of operations.

f. To meet the Technical Specification (reference 10.3.1) requirecent for having at least one member of the operating shift " qualified to implement radiation protect ion procedures", at least one, and normally three Radiation / Chemistry Technicians will be assigned to each shif t.

g. To provide for round-the-clock maintenance coverage, a maintenance crew is assigned to each shift. This crew 4-52 Am. 4 1457 021

typically consists of a Maintenance Foreman and several craft personnel providing espability in the mechanical, electrical, and instrumentation and controls disciplines.

h. The Site Security Force provides, on a rount.-the-clock basis, security services (e.g. access control, surveillance, response, etc.) in accordance with the Security Plan and procedures .

4.5.1.2 Normal Technical Suoport Organization The TMI Generation Group normal tec'anical support organization is shown on Figure 11. As can be seen in a review of the organization chart, the normal technical support organization provides the capabilities described below:

1. The Environmental, Health and Safety Division provides the administrative and technical direct ion for licensing and environmental monitoring programs. The Director of this Division reports to the head of the TMI Generarion Group who will normally be a Vice President of Met-Ed/GPU.

2. The Director of keliability Engineering p svides adminis-trative and technical direction to the Manager of Quality Assurance and the Manager of Systems Lab. The Manager of Quality Assurance is responsible for the corporate and operational quality assurance programs. The Manager of Systems Lab is responsible for providing chemistry, metal-lurgy, and other such laboratory services. The Director 4-53 Am. 4 1k7 h1

of this division also reports to the head of the TMI Generat ion Group.

3. The Director of Maintenance has overall responsibility for construction projects and corrective maintenance activities at the TMI Nuclear Station. He will report to the head of the TMI Generation Group. The Manager of Construction is directly responsible for the manage-

             =ent and coordination of design and construction activities of Met-Ed and contractors during construction projects.

The Superintendent of Maintenance, having overall respon-sibility for corrective maintenance at the TMI Nuclear Station, will work closely with the Unit 1 Superintendent who has overall responsibility for the prevent ive main-tenance program. The Unit i Supervisor of Correct ive Maintenance, who is stationed along with his staff at the site, will coordinate corrective maintenance activities with the Unit 1 Supervisor of Preventive Maintenance. 4 Engineering capability within the TMI Generation Group technical support organization is overseen by the Director of Technical Functions. He reports to the head of the TMI Generation Group. 4.5.J.3 Onsite Emergency Organization The " ice President of Nuclear Operat ions will be responsible for assigning TMI Nuc19:r Station personnel to positions in the onsite emergency organization which is shown as Figure 12. This responsibility includes the assignment of key TMI Nuclear 4-54 Am. 4 1457 023

Station management and supervisory personnel to .ne Duty Section and to the position of Emergency Director. The Unit 1 Super-intendent and the Emergency Planning Coordinator will assist him in making these assignments. The Duty Section assignments will be made and posted in writing (eg. memo, duty roster, procedure, etc.) to ensure that full 24 hour per day, 7 day per week coverage is provided. In addition, provisions will be made to ensure that alternates to key positions are pre-designated. Additional information regarding the onsite emergency organ-ization is provided below:

1. Direction and Coordination The Vice President of Nuclear Operations has overall respon-sibility for site administrat ion as well as for direct ing and coordinating the activities at TMI to ensure that it funct ions in a safe, reliable and efficient manner. He or his designated alternate, shall maintain these respon-sibilities in an emergency situation by assuming the position of Emergency Support Director.

The Unit 1 Superintendent is responsible to the Vice President of Nuclear Operat ions for the safe, reliable, and efficient operation of the plant in conformance with the Operating License. He, or his designated alternate, shall maintain these responsibilities in an emergency situation. As mentioned in subsection 5.1.1 2.a above, the Shif t Supervisor will, in the event an emergency exists, assume 4-55 g, 4 1457 024

the respensibilities of the Emergency Director. In addit ion, he will notify the Diity Sect ion Superintendent , who will relieve the Shift Supervisor of Emergency Director responsibilities upon his arrival in the Control Room. The Emergency Director, following notification c.,f an existing or potential emergency, will respond to the emergency as described in Section 6.1 below. The Emergency Director will be responsible for final assessments of emergency situations, especially where the emergency presents a real or potential hazard to of fsite persons or property. Regardless of exist ing plans, the judgement of the Emergency Director will be extremely important in assessing emergency situations and in taking appropriate protect ive and corrective actions. As such, he will implement the TMI Emergency Plan through the use of specific Emergency Plan Implementing Procedures, activate necessary and/or required portions cf the emergency organization and, as appropriate:

a. Establish methods, including communicat ions with the onsite Technical Support Center and the Offsite Emer-gency Support Center, to ensure that he will be kept informed of the status of the emergency.

b. Provide direction and support to the Shift Supervisor.

c. Provide liaison and communications with the County, State, and Federal governments, and ensure that noti-fication and reports to these agencies are made in a timely manner.

4-56 Am. 4 1457 025

d. Communicate with the of fsite emergency support organi-zation.

e. Request assistance from onsite and of fsite personnel, organizations, and agencies.

f. Organize, direct, and coordinate emergency teams.

g. Interpret radiological data obtained in terms of both real-time measurements and, to the extent possible, projected radiological exposures.

h. Ensure that adequate protective me asures are taken by personnel performing emergency ef forts. This includes those individuals assigned to the Emergency Control, Technical Support, and Operations Support Centers.

i. Ensure that accurate exposure records are maintained.

j. Review and evaluate updated information and data,

k. Relay significant information and data to casite and of fsite organizations, agencies, and response teams.

1. Determine the necessity for onsite evacuation.

2. Plant Staff Emergency Assignments

a. Communicator The Communicator will report to the Emergency Director.

He will function as liaison between the Emergency Director, offsite organizations and agencies, ons;*e organizations and agencies, and the onsite emergency organization (i.e. Technical Support Center Coordinator, Group Leader - Technical Support, NRC-Bethesda, and the 4-57 Am. 4 1457 026

Babcock and 'a'ilcox Company) . As required, the Communi-cator will provide, using available i .uipment, reliable and accurate communications in accordance with the appropriate Emergency Plan Implementing Procedures. In addition, he is responsible for maintaining records of outgoing and incoming communications. He will have designated assistants that will be utilized in an emergency as necessary.

b. Technical Support The Technical Support Center Coordinator and his staf f of engineers will report to the Emergency Director.

The technical suppcrt petsonnel will analyze current and projected plant status and, through close communi-cations (via the Communicator) with the Emergency Director, provide technical support and recommendations regarding emergency actions. In addition, the Technical support Center Cocedinator will, as necessary, provide a direct interface with the Group Leader - Technical / Support. More specifically, they will: (1) Analyze mechauical, electrical, and instrument and control problems; determine alternate solutions, design and coordinate the installation of short-term modif # cations . (2) Analyze therechydraulic and thermodynamic problems and develop problem resolutions. 1457 027 4-58 Am. 4

(3) Assist in the development of Emergency Procedures, Operating Procedures, etc. as necessary for conduct ing emergency operat ions. (4) Analyze conditions and develop guidance for the Emergency Directer end operations personnel on protection of the core. (5) Resolve questions concerning Operating License requirements with NRC representatives,

c. Plant Operations (1) The Operations Coordinator is responsible for coordinating operations and maintenance act ivi-ties through the Shift Supervisor and the Emer-gency Maintenance Coordinator. The Operat ions Coordinator may not relieve the Shift Supervisor of or specifically a rect plant operations unless he is a licensed Senior Reactor Operator. The Operat ions Coordinator will report to the Emer-gency Director.

(2) The TMI Unit 1 Operc . tons Shi f t , under the direct ion of the Shif t Supervisor, is re.ponsible for the safe and proper operation of the plant at all times. Therefore, the operations shift will respond to all abnormal and emergency situations and take act ion as necessary to improve and/or terminate any accident. 4-59 Am. 4 1457 028

The shift organization will be self-rel'snt for a sufficient period of time to allow for the notification of required personnel and for them to assemble and integrate into the emergency organization. To ensure the shif t can respond and function in an emergency, the Shift Super-visor is responsible for the initial assess-ment and evaluation of the situation and will initiate the necessary immediate actions to limit the consequences of the accident and ' ring it under control. In addition, he will assume and carry out the responsibilities of the Emergency Director until relieved by the member of the TMI Staff assigned this duty. The Shif t Technical Engineer, as discussed in subsection 5.1.1 2.e. above, will advise the Shift Supervisor on activities that impact the safe and proper operation of the plant. The shift organization personnel are familiar vith the operation of plant systems and the location and use of emergency equipment. Some members of each shift are trained in fire fighting, first aid, and the use of radiation monitoring equipment. In addition, a Radiation / Chemistry technician assigned to each shif t to provide 24 hour per day, 7 day per week coverage. 4-60 Am. 4 1457 029

(3) The Emergency Maintenance Coordinator is respon-sible for directing the activities of the Emergency Repair Team (s) and for coordinating those activities with the Shif t Supervisor through the Operations Coordinator. The Emergency Repair Team will receive health physics support , if necessary, through the Operations Support Center Coordinator.

d. Security and Personnel Accountability The TMI Site Security Force will operate by tne require-ments established in the Security Plan and Procedures.

The TMI Site Security Force will report to the Security Coordinator in emergency situations. The Security Coor-dinator, in turn, shall report to the Emergency Director. The Security Force will always respond and provide assistance as required for security violations. The responsibilities associated with persennel accoun-tability will be assumed by Security. Provisions have been made in the Security Plan for admitting emergency vehicles wher. Security is notified, in advance, by the Shift Supervisor or Emergency Director. In addition, Security will, as appropriate, direct such vehicles to the proper location. Requirements 2or providing advance notification to Security will be detailed in the appropriate Emergency Plan Implement ing and Security Procedures. 57 030 4 61 Am. 4

e. Radiological Assessment The Radiological Assessment Coordinator is responsible for direct ing the Operations Support Center Coordinator and the Ra.diological Analysis Support Engineers.

In addition he is responsible for coordinating the activities of several emergency response teams that are within his part of the onsite emergency response organization. As required, he shall direct the Opera-tions Support Center Coordinator to dispatch Offsite and/or Onsite Radiological Monitoring Teams which will report directly back to him. He shall coordinate radiological assessment act ivit ie s , review results, and report findings and make recommendations to the Emergency Director. In addition, he shall interface with the Environmental Assessment Coordinator on radiological and environmental matters. (1) The Radiological Analysis Support Engineers shall perform dose project ion calculations, source term calculations, and other such calculations or determinations that may be necessary to assess radiological hazards and to minimize personnel exposure. (2) The Operat ions Support Center Coordinator shall be responsible for and direct the activities of personnel reporting to the Operations Support i457 031 4-62 Am. 4

Center. In addition, he will also direct and coordinate the activities of the Health Physics and Chemistry Coordinators. As mentioned above, he will dispatch Offsite and/or Onsite Radio-logical Monitoring Teams as directed by the Radiological Assessment Coordinator.

f. Firefighting Specific pe sonnel on each shift (Shift Fire Brigade) are trained in firefighting to ensure such capability will be available 24 hours per day, 7 days per week.

The Fire Brigade, under the direct ion of the Shif t Supervisor or another individual designated by him,

, shall respond to all fire alarms and report to the location of the fire with assigned equipment. Curing the normal work week, additional qualified firefighting personnel will, as necessary, be obtained from the normal onsite organization. Assistance will be requested from local fire departments as deemed necessary by the Shif t Supervisor.

g. First Aid and Rescue Medical emergencies and rescue operations will be the responsibility of the First Aid and Rescue Team.

Specific personnel on each shif t are trained in first aid techniques to ensure such assistance will be available 24 hours per day, 7 days per week. Assis-tance will be requested frem outside medical support 4-63 un. 4 1457 032

personnel or organi:ations as deemed necessary by the Emergency Dire.9 tor.

h. Radiological Monitoring Prior to the activation of the entire onsite emergency organization, the Emergency Director may dispatch Of fsite and/or Onsite Radiological Monitoring Teams as well as other personnel to perfore radiological monitoring functions. After the onsite emergency organization is activiated, the Operations Support Center Coordinator will control the dispatching of all radiological monitoring funct ions. However, af ter the Of fsite and Onsite Radiological Monitoring Teams are dispatched, they shall report directly to the Radiological Assessment Coordinator. The various radiological monitors and monitoring teams are responsible for performing radiation / contamination surveys, for radiological monitoring, and for assisting in decontamination activities as assigned.

i. Repair and Damage Control The Emergency Repair Team (s), under the direction of the Emergency Maintenance Coordinator, provide for the assessment of equipment damige and will effect emergency repairs as required for operations. In addition, operations personnel assigned to these teams will be able to perform e=ergency operations as well.

1457 033 4-64 Am. 4

4.5.1.4 Of fsite Emereenev Sup port Orcanization The of fsite emergency support organization, shown as Figure 13, will provide technical and logistics support in the event of a serious or potentially serious emergency. This organization

       ----~ will be staf fed by personnel from the normal station organi-zation, the normal technical support organization, and consul-tants. In general, the responsibilities of the offsite emergency support organization are to:

o Provide liaison and communications with the Nuclear Regulatory Commission and the appropriate State and county agencies. o Provide public relations and make news releases. o Provide for envienomental monitoring and assessment in support of the onsite emergency organization. o Provide security support. o Support the onsite emergency organization in engineering and technical matters. o Coordinate the restoration and/or operation of all genera-tion, transmission and distribut ion facilities. o Procure and dispatch transportat ion equipment , and services. o Purchase materials, equipment, and services necessitated by the emergency. o Provide assistance for reentry operations and post-accident planning. o Assign post-accident investigation and review responsi-bilities. 1457 034 4-65 Am. 4

- ~ ~ _ _ ._

l. The Emergerac y Support Director and his Assistast will be responsible for activating and directing the off ite emergancy sv ' port organization and ensuring that the functional groups provide a coordinated response in support-of the onsite emergency organization. The Emergency Support Director will serve as the senior management representative at or in the vicinity of the TMI site. As such, during emergency operations, the Emergency Support .

Director may provide advice and guidance to the Ecergency Director: the Emergency Director, however, will maintain overall responsibility for the operation and control of the plant. As emergency situations tend to stabilize, the Emergency Support Director may relieve the Emergency Director of more and more accident management responsi - bilities. This will provide a controlled means of shifting to a recovery organization should that type of organiza'.$.nal arrangement be deemed necessary.

2. The Emergency Support Communicator and his staf f of assis-tants will be responsible for the operation of the com-munic tions systems at the Offsite Emergency Support Center and the coordination of requests for outside assistance.

Their dut ies include , but are not necessarily limited to:

a. The setup and operation of primary communications systems.

1457 05-3 4-66 Am. 4

b. The setup and operation of backup communicat ions systems.

c. Maint aining records of communicat ions.

d. Maintaining the Emergency Support Cer.ter status board up-to-date.

e. Coordinating the procurement of outside resources (eg.

technical assistance, manpower, equipment, etc.) with the Group Leader of Administrative Support.

3. The Public Af fairs representative shall report to the Emergency Support Director in a staf f capacity. His toecific responsibilities are detailed in the Emergency Public Information Plan which is attached to this Plan as Appendix B. Briefly, however, his primary responsi-bilities will be to maintain liaison with Met-Ed/GPU management in Read?'g and keep them informed of all onsite developments and prepare news releases. To fulfill these responsibilities, he is responsible to the Emergency Support Director for the preparation of technically accurate information for media release.

4. The Group l eader-Maintenance Support and his staff will be responsible for coordinating the allocation of Company (i.e. Met-Ed) and non-Company maintenance equipecnt and manpower to support onsite operations and maintenance ac t ivi t ies .

1457 036

5. The Group Leader-Technical Support and his staf f will be respone'ble for providing technical analysis, evaluation, 4-67

_- . ~ _ _ _ . _ . . - and recommendat ions to the Etergency Support Director and the onsite Technical Support Center Coordinator and his staff with respect to plant conditions, reactor core status, and subsequent plant operations. More specifically they will:

a. Analyze mechanical, electrical, and instrument and control problems; determine alternate solutions, design and assist in the coordination of the installation of short-term modifications.

b. Analyze thermohydraulic and thermodynamic problems and develop problem resolutions.

c. Assist in the development of Emergency Procedures, Operating Procedures, etc. as necessary for conducting emergency operations.

d. Analyze conditions and develop guidance for the Emergency Support Director, the Emergency Director, and operations personnel on protection of the core.

e. Resolve questions concerning Operating License require-ments with NRC representatives.

6. The Group Leader-Sec 2rity Support will be responsible for the overall security program required to support emergency operations. This includes the coordination of permanent and temporary security forces and the clearance and badging of emergency response personnel requiring site access. M Group Leader-Security Support will provide a 7 ember of the Security Force to the Group Leader-Administrative Support 4-68 Am. 4 1457 037

to assist him in the processing and clearnace of emergency response personnel requiring site access. . TI- 1roup Leader-Health Physics / Chemistry Support and his staff will be responsible for all aspects of health physics, chemistry, and environmental assessment support for the onsite emergency organization. Such support may include, but is not limited to:

a. Overall assessment of the impact of liquid and gaseous effluents with respect to Technical Specifications (reference 10.13.2) and protect ive act ion guices.

b. Determination of in plant sampling requirements based on plant conditions.

c. Coordination of the use of laboratory instrumentation, sample analysis, sample storage, and the interpre-tation and dissemination of analysis results.

d. Providing a member of the Health Physics Department to the Group Leader-Administrative Support to assist him in the processing (including issuing personnel monitoring devices) of emergency response personnel requiring site access.

e. Personnel dosimetry (internal and external) and emergency TLD controls and documentation.

f. Whole body counting and evaluation.

g. Identification and co< rdination of required equipment and manpower resources.

h. Assessment of solid waste problems. T457 0 %

4-69 Am. 4

8. The Group Leader-Administrative Support and his staff will be responsible for administrative and logistics functions required to support the entire offsite and onsite emergency organizations. The types of support services that might be required include, but are not necessarily limited to:

a. General Administration (1) Word processing (2) Typing pool (3) Reproduction

b. Trans port at ion (1) Helicopter services (2) Vans, busses, automobiles, shuttle service

c. Personnel Administration and Accommodations (1) Personnel processing (a) Registration (b) Indoctrination and training (c) TLD issue (d) Security badging (2) Lodging, food

d. Outside Plant Support (1) Trailer set-up (2) Janitorial service (3) Telephones

e. Commissary (1) Temporary facilities and meal delivery

f. Safety (1) Industrial safety 1457 039 4-70 Am. 4

(2) First Aid (3) Training

g. Human Resources (1) Manpower (2) Labor relations (3) Payroll (4) Clerical Support 4.5.1.5 Long-Term Recoverv Organization In those cases where post-accident conditions indicate that recovery operations will be either complicated or will extend over a reutively long period of time, Met-Ed will shif t from the emergen:y res h we organizations (i.e. onsite and offsite support) to a long-term recovery organization. The organization itself will, of course be dependent upon the nature of the accident, post-accident conditions (i.e. plant conditions, radiation / contamination levels, etc.) and other factors to be determined at the time. Prior to initiating recovery operations, which are further discussed in Section 9.0 below, a specifi:

long-term recovery organization will be defined. A typical long-term recovery organization is shown on Figure 14 and is described below:

1. The President of GPU is responsible for selecting the senior parsonnel to fill the key position in the long-term recovery organization and for implementation and coor-dination of recovery operations.

4-71 Am. 4 1457 040

2. The Recovery Operations Manager is responsible for the overall technical aspects of the recovery operation. This includes overseeing the operations of the various functional groups and ensuring that all activities, proposed courses of action, and contingency plans receive proper analysis and coordination.

3. The Public/ Government Af fairs Group is responsible for coordinating the exchange of information with public and go"ernmental agencies and includes legal counsel as a par of its organization.

4 The Administration and Logistics Group is resconsible for providing the necessary administrative / logistics require-ment s such as communications , manpower, transportat ion, commissary arrangements, accommodations, clerical support, and temporary office space and equipment.

5. The Technical Working Group includes the heads of each of the technical groups discussed in the following paragraphs.

In addition, representatives from the nuclear steam supply systems supplier, the architect engineer, consultants, the Nuclear Regulatory Commission, and other individuals shall be included in the Technical Working Group as appropriate.

6. The Task Management / Scheduling Group r --iorities; develps pland and schedules; coordinates and monitors the status of tasks; and reports the work progress of all the technical groups. In addition, the group provides liaison with the Nuclear Regulatory Commission.

4-72 Am. 4 1457 041

7. The Technical Support Group is responsible for providing engineering support, technical plancing and analysis, procedure support, contro'. room techn' cal support, data reduction and management, and support relating to licensing requirements.

8. The Plant Operations Group consists of the plant staff with substantial augmentation from other organizations.

This group is responsible for performing all plant opera-tions and maintenance activities, limiting and controlling personnel er.posures, in plant hea .th 1 physics management , terminating or minimizing of fsite releases, stabilizing plant conditions, and restoring the plant's ability to function nomally and respond to any further emergencies.

9. The Waste Mangement Group is responsible for safely and ef fect ively managing the quant ities of radioactive gases, liquids, and solids that might exist during the initial phasss cf recovery. Subsequently, this group is respon-sible for the development and implementation of short and long-term plans to manage and process contaminated solids, liquids, and gases; quantifying the degree of contamir.ation of buildings and systems; and the estab-lishment of processing priorities based on plant needs.

10. The Plant Modifications Group is responsible for providing the engineering, design, materials and construction neces-sary to completa required modificatic as to plant systems, equipment, and structures.

1457 042 4-73

11. The Industry Advisory Group is designed to function in parall.el with the other technical groups and is not intended to be part of the implementation structure. The purpose of this technical group it to objectively look into potential problems, maintain a current awareness of the porceived plant and reactor core status, and provide independent assessment based on experience and judgement rather than detailed engineering review and calculations.

4.5.2 kgcal Services Support The natrre of an emergency may require augmenting the onsite emergency organization therefore, it may become necessary to request and utilize assistance furnished by local personnel, organizations, and agencies. Since it is essential that support from local law enforcement agencies, fire departments, hospitals, physicians, and ambulance services be available on relatively short notice, agreements, cope of which have been included in Appendix C, have been made with the following personnel, organization, and agencies: 4.5.2.1 Medical Support Organizations and Personnel

1. Hershey Medical Center

2. Physicians

3. Radiation Management Corporation 4 Bainbridge Fire Company (ambulance service)

5. Liberty Fire Company (ambulance service)

6. Londonderry Township Fire Company (ambulance service)

7. Rescue Hose Company No. 3 1457 043 4-74 Am. 4

4.5.2.2 Firefichtinz Organizations

1. Bainbridge Fire Company

2. Liberty Fire Company

3. Londonderry Township Fire Company

4. Rescue Hose Company No. 3 4.5.2.3 Law Enforcement Agencies

1. Pennsylvania State Police

2. Middletown Police Department 4.5.2.4 Miscellaneous Suoport

1. Consolidated Rail Corporation 4.5.3 Ceordination with Government Agencies Metropolitan Edison Company has and will continue to work closely with State, County, and federal agencies in coordinating emergency planning activities for the Emergency Planning Zone in order to ensure the health and safety of the general public. As a part of this coordination, each participating agency has been assigned specific responsibilities and authority for both emergency planning and emergency response. Also as a part of this combined ef fort , specific emergency-related not ificat ion and information repotting requirements between Met-Ed and the various part icipating agencies have been defined. Additional reporting require-ments, which are specified in reference 10.14 and in Section 6.9 of reference 10.13.1, will also be met. A brief description of the key elements of each of the participating State, county, and federal agencies is provided in the following subsect ions. Additional information pertaining to emergency related not ificat ion requirement s that act ivate :ne emergency 4-75 Am. 4 1457 044

response organizations and the subsequent information reporting require-ments is provided in Sect ion 6.1. 4.5.3.1 State Agencies The State Council of Civil Defense was established under the Commonwealth of Pennsylvania, State Council of Civil Defense Act of 1951. The State Council of Civil Defense has, in general, been responsible for providing emergency services in the State. In order to plan for emergencies in a manner that compliments federal guidelines and implements State directives that are related to emergency operations, the Conmonwealth of Pennsylvania Disaster Operations Plan was prepared and issued under the authority of, and in sccordance with the provisions of the State Council of Civil Defense Act of 1951. Annex E, Emergency Nuclear Incidents (Fixed Nuclear Fac ilit ies ), to the Disaster Operations Plan provides additional guidelance to the State, county, local, and federal agencies and nuclear power plant facilities in the development and implementation of emergency plans associated with radiological emergencies. Annex E is attached to this Plan as Appendix D.

1. State Council of Civil Defense The Governor, by law is the Chief Executive Officer of the Commonwealth of Pennsylvania and is, therefore, responsible for the safety and well-being of all citizens within the State. The Governor has accepted this responsi-bility by heading the State Council of Civil Defense under the chairmanshin of the Lieutenant Governor.

                                   ~

i457 045 4-76 Am. 4

The State Disaster Operations Plan states that, "As the legally authorized and formally designated State disaster agency, the State Council of Civil Defense exercises overall policy guidance, direction and control of emergency and major disaster operations within the Commonwealth. To farnish assistance in the discharge of its legally assigned responsibilities, the Governor has appointed a Director of Civil Defense and the State "ouncil employs a full-time administrative and emergency operational staff. In emergency planning and operations, the State Council of Civil De'ense is supported by all agencies of State Government." Specific responsibilities assigned to the State Council of Civil Defense and other supporting State agencies are defined in Annex A to the State's Disaster Operations Plan.

2. Pennsylvania Emergency Management Agency The Pennsylvania Emergency Management Agency (PEMA) has been assigned as the lead State agency for the coordination of Radiolological Emergency Response Plans prepared by the State, County, and Local governments and the TMI Emergency Plan prepared by Met-Ed. In addition, should a radiological emergency occur at the TMI Nuclear Station that requires the implementation of State, county, and local government Radiologf. cal Emergency Response Plans, the State agency through vnich the Governor and the State Council of Civil Defense will exercise coordination / control will 4-77 Am. 4 1457 046

be PEMA. However, as in all emergencies, the Governor retains direct ional control . The Director of PEMA haa the overall responsibility for the agency's operation. To ensure that PEMA can provide 24 hour per day, 7 day per week response to emerteacies, the State Emergency Operations Center will either be manned by a Duty Officer or he will be on call. Specific responsibilities assigned to PEMA are defined in Section IX of Annex E to the State's Disaster Operations Plan which is attached as Appendix D. To ensure that PEMA can adequately respond to, and cope with a radiological emergency and provide the necessary interfaces and support, the Department of Environmental Resources, Bureau of Radiation Protect ion will, in an advisory capacity, work with PEMA.

3. Department of Environmental Resources The Department of Environmental Resources (DER), under the administration and technical direction of the Director, DER, is generally responsible for gathering and evaluating technical information and for supplying such information and technical advice and recom=endations to PEMA and the State Council of Civil Defense. Specific responsibilities assigned to the DER are defined in Annex A to the State's Disaster Operations Plan.

Within the DER, the Bureau of Radiat ion Protection (BRP) has been delegated DER's responsibilities as they apply to 4-78 Am. 4 1457 047

radiological emergencies. The Director, BRP, is respon-sible for meeting these delegated responsibilities, however, the Director, DER, will maintain overall responsibility. Specific responsibilities assigned to the DER /BRP that are appropriate to radiological emergencies are defined in Section IX of Annex E to the State's Disaster Operations Plan which is attached as Appendix D. To provide for emergency response capability, the BRP has made provisions for 24 hour, 7 day per week interface with PEMA. 4.5.3.2 County Agencies Section 7 of the State Council of Civil Defense Act of 1951 states that "Each political subdivision of this State is hereby authorized and directed to establish a local organization for civil defense in accordance with the State Civil Defense plan and program. Each local organization for civil defense s'aall have a Director who shall be appointed by the Governor upon the recommendation of the executive officer or governing body of the political subdivision. The Director shall be responsible for the organization, administration and operation of such local organization for civil defense. subject to the direction and control of such executive of ficer or governing body." Therefore, each County and Local Civil Defense Director in the State is responsible for establishing a civil defense organization within their respect ive jurisdiction, devel-4-79 Am. 4 1457 048

oping plans and preparing for emergency operations in conformity with the State's Disaster Operations Plan and the State Council of Civil Defense Act of 1951. With respect to the Emergency Planning Zone, the 5 counties identified below have prepared emergency plans that are coordinated not only with the State's Disaster Operations Plan but with the TMI Emergency Plan as well. Local covernment plans are either included directly with their respect ive County's plan or are maintained as separate, but coordinated documents.

1. Dauphin County The Dauphin County Action and Response Plan for Emergency Personnel and Citizens was prepared by the Dauphin County Office of Emergency Preparedness. This plan is attached as Appendix E.

To provide adequate response capability, the County Emergency Operations Center is the location of the County dispatch for police, fire, and rescue services and is manned by a dispatcher 24 hours per day, 7 days per week.

2. York County The York County Evacuation Plan for the Three Mile Island Nuclear Power Plant, attached to this Plan as Appendix F, was prepared under direction of the County Commissioners and the County Civil Defense Director.

1457 049 4-80 Am. 4

The York County Emergency Operations Center is the location of the County dispatcher for police, fire, and rescue services and is manned by a dispatcher on a 24 hour per day, 7 day per week basis.

3. Lancaster County The Lancaster County Emergency Evacuation Plan was prepared by the County Emergency Management Agency under the direct ion of the Lancaster County Commissioners. This plan is attached to this document as Appendix G.

The Lancaster County Emergency Mangement Agency has made provisions for 24 hours per day, 7 day per week emergency response coverage.

4. Cumberland County The Cumberland County Evacuation Plan 79-1 (For Response to a Nuclear Incident at Three Mile Island Nuclear Station) was prepared under the direction of the Cumberland County Commis-sioners and Director of Emergen:y Preparedness. A copy of this plan is attached as Appendix H.

The Cumberland County Emergency Operations Center is the location of the County dispatcher for police, fire, and rescue services and is manned by a dispatcher on a 24 hour per day, 7 day per week basis.

5. Lebanon County The Lebanon County Emergency Operations "lan, a copy of which 4-81 Am. 4 1457 050

is attached as Appendix I, was prepared by the Cumberland County Emergency Management Agency and approved by the County Commi ss ione rs . 4.5.3.3 Federal Agencies Should an emergency situation or accident occur at the TMI site, notification and/or reports and/or requests for assistance may be made to various federal agencies and organizations. Specific details for notifying and making reports to these agencies, as well as requesting and obtaining assi.:tance, will be provided in the Emergency Plan Implementing Procedures. The f,110 wing agencies may, as the situation warrant s, require not ification and/or information reports, or provide assistance if requested.

1. Nuclear Regulatory Commission The Nuclear Regulator- Commission (NRC) requires notification and reports as specified in reference 10.14 and Section 6.9 of reference 10.13.1. In addition, following notification by Met-Ed, it is expected that PEMA will also notify the NRC of emergencies at the TMI Nuclear Station. As situations warrant, the NRC Regional Of fice will notify the Department of Energy.

During a radiation emergency, the primary role of the NRC is that of conduciing investigative activities associated with the incident and verifying that emergency plans have been implemented and proper agencies notified. Should NRC personnel be dispatched to the scene they wil', as 4-82 Am. 4 1457 051

ne eded , ass ist , in .oordination with Radiologicai Assistance Teams provided by the Department of Energy, in providing to State and local agencies advisory assistance associated with investigating and assessing radiological hazards to . _ _ . . the public.. . _ _. To ensure reports can always be made, the "egion I NRC Of ficer of Inspect ion and Enforcement (I&E) is equipped with a 24 hour emergency telephone number and has assigned duty officers. The Region I I&E Office is located in King of Prussia, Pennsylvania. A lettter of understanding is provided in Appendix C.

2. Department of Energy The Department of Energy (DOE) has established several ragional coordinating offices for the purpose of providing radiological emergency assistance to State, county, and local governments; nuclear power plant facilities; and other federal agencies.

To ensure 24 hour, 7 day per week emergency response capability, the DOE has established a continuously manned Security Operations Center in each of their designated regions. The Security Operations Center for the region in which the TMI Nuclear Station is located in Upton, New York, and the Brookhaven Office. A letter of understanding is provided in Appendix C. 4-83 Am. 4 1457 052

3. U.S. Coast Guard During a radiation emergency which could involve exposure to of fsite personnel, the U.S. Coast Guard will provide assist 6nce by maintaining traffic control on the Susque-hanna River. By authority of the Pennsylvania Emergency Management Agency, the U.S. Coast Guard, as assisted by the Coast Guard Auxiliary, may also be used to keep unauthorized personnel from entering Three Mile Island and adjacent islands until emergency conditions are con-trolled. Assistance may be obtained by calling the 24 hour emergency telephone. A letter of understanding is provided in Appendix C.

4 Department of the Army The 56th Ordinance Detachment of the Department of the Army will provide Explosive Ordinance Disposal personnel in response to requests for assistance during bomb threats. Additional details of this arrangement are specified in a letter agreement which is provided in Appendix C.

5. Federal Aviation Adninistration TBD - A letter of agreement with the Federal Aviation Administrat ion is being pursued.

6. National Weather Service TBD - A letter of agreement with the National Weather Service is being pursued.

9 , .. 4-84 Am. 4 1457 053

. ,_n _ . - - . . . . - - 4.6 EMERGENCY MEASURES This Section identifies the specific measures that will be taken for each class of emergency defined in Section 4.0 of this Plan. The logic presented in this Section is used as the basis for detailed Emergency Plan Implementing Procedures which defire the emergency actions to be taken for each emergency class. Emergency measures all begin with (1) the recognition and declaration of an emergency class, (2) notifi-cat ion of the applicable agencies for each emergency class, and (3) mobilization of the appropriate portions of the emergency organi-

             =ation. The additional measures are crganized into assecament actions, correct ive act ions , protective actions and aid to af fected pers;anel and are described in the sect ions below for each emergency cla .8 4.6.1 Actization of Emergency Organizations Meeting or exceeding a predetermined value or condition specified as an emergency act ion level in an Emergency Plan Implementing Procedure shall require the implementation of that precedure. Specific emergency action levels for each emergency class (i.e. Unusual Event , Alert, Site Emergency, and General Emergency) are defined in Section 4.1 above. The Sh i f t Supervisor, in implementing the Plan will initially classify the emergency, ensure that required notificat ions are made, and not ify the Duty Sectior. Superintendent or his designated alternate. The Duty Section Superintendent will, working cln=ely with the Shif'; Supervisor, perform an overall assessment of the emergency in order to determine its most appropriate classification and, based on this determination, activate larger protions of, or the ent ire emergency organization. A more detailed discussion of the methedology that is used in act ivating the emergency 4-85                                Am. 4 1457 054

organizations during each class of emergency is provided below. In addition Figures 15,16, and 17 provide a visual display of the communi-cations networks that have been planned for notification requirements, information reporting, and decision making with respect to taking pro-tective action for the public, respectively. 4.6.1.1 Shift Foreman / Control Room Operators

1. Should emergency situations (real or potential) arise, it is expected that the Control Room Operator (s) and/or the Shift Foreman will be initially made aware by alarms, instrument readings, reports, etc. The Control Room Operator (s) shall ensure that the Shift Foreman and/or the Shi f t Supervisor are immediately informed. The Shif t Foreman shall, if not already accomplished by the Control Room Operator (s), inform the Shift Supervisor immediately.

4.6.1.2 Shift Supervisor

1. The Shift Supervisor, when informed of an emergency situ-at ion, is responsible for performing the assessment of the emergency (e.g. plant systems and reactor core status, radiological conditions, etc.) in the following manner:

a. By determining what imediate actions e tst be taken (e.g. use of Emergency Procedures) to ensure the safe and proper operation of the plant. The Shift Technical Engineer will advise and assist the Shift Supervisor on matters pertaining to the safe and proper operation of the plant with regards to nuclear safety.

4-86 Am. 4 05,3 1457

2. If the situat ion requires implementation of the TMI Emergency Plan, the Shift Supervisor shall:

a. Initially classify the emergency as an Unusual Event, Alert, Site Emergency, or General Emergency as appropriate and implement the applicable Emergency Plan Implementing Procedure.

b. Not ify the followng agencies and personnel of the emerceney declaration:

(1) Dauphin County EOC (2) PEMA EOC (3) NRC Region I Offit.e of Inspection and Enforcement (4) NRC Office - Betheada, Maryland (5) Duty Sect ion Superintendent

3. Due to the numerous responsibilities assigned to the Shift Supervisor at the onset of an emergency, he shall prioritize his actions to (1) ensure the safe operation of the plant ,

(2) ensure that immediate notification requirements are met, (3) dispatch, in the event of' radiological emergencies, Radiological Monitoring Teams, and (4) perform additional emergency actions as time aad conditions permit. 4.6.1.3 Duty Section Superintendent

1. The Duty Sect ion Superintendent , upon being informed of an emergency declaration by the Shift Supervisor shall:

a. Evaluate the information, data, and methods ased by the Shift Supervisor in his determination in 4-87 Am. 4 1457 056

order to ensure that the proper emergency classifica-tion has been made.

b. Determine to what extent the offsite and onsite emergency organizations shall be activated within the following guidelines:

(1) For an Unusual Event, all or part of the Duty Sect ion shall be act iviated. (2) For an Alert, the Duty Sect ion shall be activated along with all or portions of the onsite emergency organization. (3) For a Site Emergency, the Duty Sect ion and th= ent ire onsite emergency organization shall be act ivat ed . A major portion, if not all of the offsite emergency organization shall also be activated. (4) For a General Emergency, the entire onsite and ot.' site emergency organizations shall be acti-vated.

c. Inform the support member of the Duty Section of the emergency and direct him to act ivate the of fsite and/or onsite emergency organizations as determined immediately above.

d. Report to the Emergency Control Center and assume the position of Emergency Director in the onsite emergency organization.

4-88 Am. 4 1457 057

NOTE The Shift Supervisor shall assume the responsibilities assigned to the Emergency Director until properly relieved by the Duty Section Superintendent. 4.6.1.4 Emergency Director

1. The Emergency Director shall locate himself in the Emergency Control Center and assume the responsibilities specified in paragraph 1 of subsect ion 5.1.3 above.

4.6.1.5 Dauphin County

1. The dispatcher at the Dauphin County EOC shall notify the Dauphin County Civil Defense Director or his designated alternate.

2. The Dauphin County Civil Defens, shall, through area and local civil defense personnel and other means, notify local municipalities with priority favoring those nearest TMI.

4.6.1.6 Pennsylvania Emergency Management Agency

1. The PEMA duty of ficer at the State EOC shall, upon receiving notification of an emergency at the TMI Nuclear Station, immediately notify the Bureau of Radiation Protection.

2. DEMA shall, in accordance with standard operat ing procedures ,

notify the following personnel, organizations, and agencies as appropriate:

a. Dauphin County Civil Defense

b. Other af fected county civil defense organizations (e.g.

York, Lancaster, Cumberland, Lebanon)

c. Other affected states (i.e. Maryland) 4-89 Am. 4 1457 058

d. Selected State agencies such as the:

(1) Department of Environmental Resources (2) Department of Agriculture (3) Depart. ment of Health (4) State Police (5) Department of Transportation

e. Selected Federal agencies such as the:

(1) Nuclear Regulatory Commission (2) Department of Enero y (3) Environmental Protect ion Agency (4) Federal Emergency Management Agency 4.6.1.7 Bureau of Radiation Protection

1. The Bureau of Radiation Protect ion employee (i.e.

Incident Manager) who receives the notificat ien from PEMA that an emergency exists at the TMI Nuclear Station shall:

a. Call the TMI Nuclear Station Emergency Control Center to:

(1) Verify the classification of the emergency (2) Cbtain information and data pertaining to the emergency.

b. Initiate activation of the BRP emergency response organization.

c. Advise the PEMA duty of ficer of the BRP's initial assessment of the emergency.

1457 059 4-90 Am. 4

4.6.2 Assessment Actions Ef fective coordination and direction of all elements of the emergency organization requires cont inuing accident assessment throughout an emergency situation. Each emergency class will invoke similar assessment methods, however, each class imposes a different magnitude of assessment effort. In the following subsections assessments actions to be taken for each emergency class are outlined. 4.6.2.1 Assessment Actions for Unusual Events The detection of an Unusual Event will arise from either exceeding a specific emergency action level for this class (see Subsect ion 4.1.1) or by an action statement in an Emergency Procedure. Detect ion of the event in either case will come as the result of alarms, instrument readings, recognit ion through experience or a combinat ion thereof. The continuing assessment act ions to be performed for this class of emergency will be in accordance with the Emergency Plan Implementing Procedure for an Unusual Event and will consist of the normal monitoring of Control Room and other plant instrumentat ion and status indications until the situation is resolved. If a fire was the reason for the declara-tion of an Unusual Event, the Fire Brigade leader who will have reported to the fire location will make continuing assessments based on his experience and report to the Emergency Director on whether offsite firefighting assistance is required. 4.6.2.2 Assessment Actions for Alerts Once an accident has been classified as an Alert by the Emergency Director, assessment act ions will be performed in accordance 4-91 Am. 4 1457 060

with the Emergency Plan Implementing Procedure for an Alert. These act ions will include:

1. Increased surveillance of in plant instrumentation.

2. If possible, the dispatching of an Emergency Repair Team to the identified problem area for confirmation and visual assessment of the problem.

3. The dispatch of onsite radiological monitoring team (s) to monitor for possible releases and provide rapid confirmation of correct accident classification.

4. Dose Assessment -- If a radiological accident is occuring, surveillance of the in plant ins t rume nt at io n necessary to obtain meteorological and radiciogical data required for calculating or estimating projected doses will commence. The dose assessment activity will continue until termination of the emergency in order that updating of initial assessments may be provided to all concerned of fsite agencies ind to the Emergency Director. Emergency Plan Implemmating Procedures are provided to allow rapid, consistent projection of doses.

4.6.2.3 Assessment Actions for Site Emer2encies The assessment act ions for the Site Emergency class are similar to the act ions for an Alert, however, due to the increased magnitude of the possible release of radioact ive material, a significantly larger assessment activity will occur. The neces-sary personnel for this assessment ef fort will be provided by 1457 061 4-92 Am. 4

mobilization of the entire onsite and offsite emergency organi-zations. Specifically,

l. An increased amount of plant instrumentation will be monitored. In particular, indications of core status (e.g. incore thermocouple readings) will occur.

2. Monitoring efforts will be greatly increased. Onsite and of fsite monitoring teams will be dispatched.

In addition to beta att gamma field measurements, the changeout of thermoluminescent dosimeters (TLD's) at frequent intervals will be done, air sampling and collection of other environmental media for assessment of material transport and deposition will be performed.

3. Dose assessment activities will be conducted more frequently, with an increased emphasis on dose pro-jection for use as a factor in determining the neces-sity for protective actions. Rad *ological and meteoro-logical instrumentation readings shall be used to project (1) dose rate at predetermined distances from the plant, and (2) an integrated dose. In reporting the dose project ion to the Emergency Director or of fsite agencies, the dose rate, dose, and the basis for the time used for the dose estimate will always be provided. As an example, an early dose assessment might include a dose rate and a dose for a 2 hour period based on an arbitrarily chosen time period. The next report, possibly 30 minutes later may report the same dose rate with a dose based en ten hours with the 4-93 1457 062 Am. 4

new time being correlated with an estimate. for repair of the damaged component. Any confirmation of dose rates Ly of fsite teams will be reflected in reporting or revising dose estimate information provided to applicable offsite agencies. Dose projections will be considered by plant personnel in relation to the EPA PAG's. Report ing of assessments to offsite officials shall include the relationship of doses to thesa guidelines. Emergency Plan Implementing Procedures shall specify report ing format as well as projection methods. In addition, forms will be provided for recording all pert inent information. 4.6.2.4 Assessment Actions for General Emergencies Assessment actions for the General Emergency Class will be the same as for the Site Emergency class with some possible shift of emphasis to greater of fsite monitoring ef forts and dose projection ef forts extending to distances f arther from the plant. Additionally, since the projected doses are likely to be much closer to EPA PAG's, greater emphasis shall be placed on the assessment of release duration. Judgements and assump-tions used for dose assessment will always be reported. 4.6.3 Corrective Actions Detailed operating procedures are available .o the operators for use during emergencies as well as normal operations. Specific Emergency Procedures are provided to assist the operators in placing the plant 1457 063 4-94 Am. 4

in a safe condition and taking the necessary supplemental corrective ac t ions . In addition, operations personnel are trained in the oper-at ion of plant systems and their associated procedures and will, there-fore, be capable of taking eppropriate corrective actions. Selected TMI Nuclear Station Staff personnel, including operations, health physics, and maintenance personnel are trained and assigned to emergency teams. These teams will be able to respond as set forth in the Emergency Plan Implementing Procedures in order to assess conditions and take any available correct ive actions. Maintenance personnel will provide the necessary crafts expertise to ef fect repair and damage control funct ions. Corrective act ions will normally be planned event s that are taken to ameliorate or terminate the emergency situation. Planned radioactive releases or corrective actions that may result in a radioactive release will be evaluated by the Emergency Director and his staff as far in advance of the event as is possible. Such events and data pertaining to the release will, if at all possible, be reported to the appropriate of fsite emergency response organizations and agencies. 4.6.4 Protective Actions Protective actions are emergency measures taken during or af ter an emergency situation that are intended to minimize or eliminate the hazard to the health and safety of the general public and/or site personnel. Such actions taken onsite are the responsibility of Metropolitan Edison Company while those taken offsite fall under the jurisdiction of the Co=monwealth of Pennsylvania and other offsite response agencies. All 4-95

                                                                      ^** '

1457 064

visitors to the site will be either escorted by an employee knowledgable as to emergency plan response or will receive training on act ions required by them during an emergency. 4.6.4.1 Protective Cover, Evacuation, Personnel Accountability During an emergency, the relocation of persons may be required in order to prevent or minimize exposure to radiation and radioactive materials. The following subsections discuss the policies applying to such situations.

1. Plant Site All persons onsite at the time of an Alert, Site or General Emergency who do not have emergency assignments (i.e.

nonessential personnel) shall be notified of the emergency class by announcement over the public address system. These persons will be instructed to report to assembly areas for accountability, monitoring and possible evacuation. Such persons will be trained in the routes to assembly areas or be escorted by an employee who is so trained. At the assembly areas, members of the emergency organization will direct and conduct accountability, monitoring and evacuation ef forts per applicable Emergency Plan Imple-menting Procedures. These procedures will provide contin-gency plans for weather, traf fic and radiological impedi-ments to evacuation (such as alternate routes) .

2. Offsite Areas The responsibility for act ions to protect persons in of fsite areas rests with the Commonwealth of Pennsyl-4-96 Am. 4 -

1457 0bD

vania and is described in detail in Annex E (attached as Appendix D) to the Commonwealth of Pennsylvania Disaster Operations Plan and implemented in conjunction with several additional county emergency plans. Met-Ed, through the TMI Emergency Director shall remain ready throughout an emergency to provide protective action recommenda-tions to State officials (see also Assessment Ac t ions , Sect ion 6.2) . The means to warn or advise persons involvec in designated a responsibility of the " Risk county" in Annex E of the State's Disaster Operations Plan. Annex E also designates the " Risk county" as responsible for the preparation and dissemination of information material on protect ive actions to the general public. 4.6.4.2 Use of Onsite Protective Equipment and Supplies The following onsite locations have been designated as emergency assembly points and areas where emergency teams will be assembled and equipped. Emergency equipment and supplies, as described below, will be maintained at these locations. The use of these assembly points and/or the emergency equipment and supplies will be in accordance with the Emergency Plan Implementing Procedures or as directed by the Emergency Director or his representatives. Such representatives may be the Shift Supervisor, or emergency team leaders.

1. Control Room / Shift Supervisor's Of fice - The Control Room and Shift Supervisor's office are designed to be habitable under accident conditions and will serve as the primary 4-97 Am. 4 1457 066

onsite Emergency Control Center (ECC). Emergency lighting, power , air 3 .tration-ventilation system, and shielding walls enable operators to remain in the t.ontrol Room to ensure that the reactor will be maintained in a safe condition. In addition, the operators will be able to evaluate planc conditions and relay pertinent information and data to appropriate onsite and of fsite personnel, organizations, and agencies during all emergencies. To ensure the opera-tions shift and other personnel assembled at the location can remain self-sufficient, emergency equipment and supplies will be stored in or near the Control Room. The exact location and the type and quantity of emergency equipment and supplies available will be specified in the Implementing Document. However, as a general guide, the following will be available:

a. High and low range dosimeters,

b. Dosimeter charger,

c. Dose rate survey meter (s),

d. Contaminat ion sarvey meter (s),

e. 'rrot ect ive clothing,

f. Respirators and/or masks,

g. Air sampler (s),

h. Portable lanterns and/or flashlights,

i. Walkie-talkie (s),

j. Site sector map and map overlays,

k. Map of environmental monitoring stations,

1. Copy of the Implementing Document, 4-98 Am. 4 1457 067

n. Station floor plan drawings and piping and instrumer.:

diagrams,

n. Resuscitator,

o. First aid kit, and

p. Miscellaneous supplies and equipment.

2. Access Control Peint - During normal operation this area serves as the access control point for personnel entering the controlled areas. Health Physics equipment will normally be on hand in this area to support dcily operat ion and maintenance act ivities. However, to ensure adequate inventory of equi ocent and supplies will be available, additional emergency equipment and supplies will be main-tained in this area. The exact location and the type and quantity of emergency equipment and supplies will be specified in the Implementing Docement. However, as a general guide, the following will be available:

a. High and low range dosimeters,

b. Dosimeter charger,

c. Dose rate survey meter (s),

d. Contaminat ion survey meter (s),

e. Protect ive clothing,

f. Respirators and/or masks,

g. Air sampler (s),

h. Neutron survey meter (s),

i. Pc; table lanterns and/or flashlights,

j. Copy of the Implemenring Document,

k. Dec ont amination chemicals .

4-99 Am. 4 1457 068

3. Process Centee The Process Center will be continuously manned by Site Security Force personnel. Emergency equip-ment and supplies will be maintained in this facility to support such tasks as reentry efforts, performing onsite radiation surveys or collecting samples. The exact location and the type and quanticy of emergency equipment and supplies will be specified in the Implementing Document.

However, as a general guide, the following will be available:

a. High and low range dosimeters,

b. Dosimeter charger,

c. Dose rate survey meter (s),

d. Contaminat ion survey meter (s),

e. Protective clothing,

f. Respirators and/or masks,

g. Air sampler (s),

h. Portable analyzer,

i. Portable alternator,

j. Portable lanterns and/or flashlights,

k. Site sector map and map overlays,

1. Map of environmental monitoring stations,

m. Walkie-talkie (s),

n. Copy of the Implement ing Document ,

o. First aid kit, and

p. Miscellaneous supplies and equipment.

4. North Assembly Area - The North Assembly Area is the auditorium in the Unit 1 Service Building. This is the assembly point for non-essential personnel in uncontrolled 4-100 Am. 4 1457 069

areas inside the protected area fence. The emergency equipment and supplies, located in lockers in the vicinity of the assembly areas, is maintained for such tasks as personnel monitoring, and readying contaminated, injured personnel for transporting to the hospital. The exact location and the type and quantity of emergency equipment and supp1'ies will be specified in the Implementing Docu-ment. However, as a genera! Iuide , the following will be available:

a. Poly sheeting,

b. Protective clothing,

c. Monitoring equipment .

4.6.4.3 contamination control Measures This section describes provisions for preventing or minimizing direct or subsequent ingestion exposure to radioactive materials deposited on the ground or other surfaces.

1. Plant Site All areas within the exclusion area are owned by Met-Ed and will, therefore, be controlled. In addition, there are no areas for producing agricultural products within the exclusion area. In plant contamination control will be controlled in accordance with approved radiation procedures.

2. Offsite Areas Offsite, it is the responsibility of the State Depart-ment of Agriculture in conjunction with the Departments 1457 070 4-101 Am. 4

of Environmental Resources and Health to issue guidance and coordinate actions to control contaminated agricul-tural products (See Annex E of the Commonwealth of Pennsylvania Disaster Operations Plan). 4.6.5 Aid to Af feoted-Personnel - This section of the plan describes the measures that will be used by the licensee to provide necessary assistance to persons injured or exposed to radiation and radioactive material. 4.6.5.1 Emergency Personnel Exposure Emergency measures may warrant the acceptance of above-normal radiation exposures (doses). Saving of life, measures to circumvent substantial exposures to population groups or even preservation of valuable installations may all be sufficient cause for above normal exposures. The following are the exposure guidelines for these emergency activities. Life - saving action - 100 Rem Corrective action - 25 Rem Personnel involved in any of the above actions must be volunteers. The Emergency Director shall authorize the above exposures and is responsible for maintaining exposures below these values. He shall, if possible, seek advice from the Radiological Assessment Coordinator. He shall assure that all measures are taken to minimize other exposures (such as internal exposure) during these act ivities. 4-102 Am. 4 1457 071

4.6.5.2 Decontamination and First Aid Decontamination materials, specialized equipment and supplies, and portable first aid kits are available in the Decontamination Area. Portable health physics instrucents for personnel monitoring and portal monitors are available at the access control point. Decontamination showers and a sink, both of which drain to the radwaste system, and an eyewash are also located in the Decontamination Area. All personnel leaving the controlled area will be monitored for contamination by use of portal monitors, and/or hand and foot countera, and/or friskers. During emergencies all personnel onsite will, as necessary, be checked for contamination. Personnel found to be contaminated will undergo decontamination by health physics personnel (or other qualified personnel as specified in Health Physics Procedures). It is preferable that personnel decontamination be performed by health physics personnel, however, other Met-Ed personnel are instructed in both decontamination and first aid procedures. Measures will be taken to prevent the spread of contaminat ion. Such measures may include isolating affected areas, placing con-taminated personnel in " clean" protective clothing before moving, and decontaminating af fected personnel, their clothing, and equipment prior to release. Emergency first aid and medical treatment will be given to injured personnel who are contaminated. Sh i f t personnel, 1457 072 4-103 Am. 4

trained in first aid, will be available onsite on a 24 hour per day basis and will assist contaminated personnel either at the scene of the accident or in tne first aid room located in the 305' elevation of the Service Building. Provisions have been made, through agreements, to ensure contaminated and injtred personnel will receive specialized medical treatment if neces-sary. The Hershey Medical Center has agreed to accept conta-minated patients for emergency medical and surgical treatment and/or observation. Trensportat ion will be arranged as provided for and discussed in subsection 6.5.3. If af fected personnel must be transported, measures will be taken to prevent the spread of contamination. Such measures may include placing af fected personnel in " clean" protective clothing or wrapping in blankets and alerting the organizations who will provide the trans port ation and treatment. If necessary, a physician will be requested to provide medical assistance onsite. Detailed instruct ions for treatment and transportation of contaminated and injured personnel will be specified in the Implementing Document. 4.6.5.3 Medical Transoortation Ambulance sarvice for the facility is provided for by letters of agreement with neighboring fire departments. See Appendix C for actual agreement letters. 4.6.5.4 Medical Treatment Arrangements for hospital and medical services for injured or contaminated / overexposed personnel are provided for by letters of 4-104 ^** 4 1457 073

                                               =   - - ~ , _ _ . -

agreement with the Hershey Medical Center, Radiation Management Corporation and local physicians. See Appendix C for actual agreement le tt ers . The services of the Radiation Management Corporation assure that personnel providing services are prepared and qualified to handle radiological emergencies. 4-103 Am. 4 1457 074

4.7 EMERCENCY FACILITIES AND EQUIPMENT This section describes the equipment and facilities that are utilized to: o Assess the extent of accident hazards. o Mobilize the resources required to mitigate the consequences of an accident. o Provide protect ion to plant personnel. o Support accident mitigation operations. o Provide immediate care for injured personnel. o Affect damage control. A diagram identifying the major emergency facilities and their general location relative to each other is attached as Figure 18. Many of the Met-Ed facilities and much of the equipment is normally used for routine plant operations. Other items are reserved for use only on an "as needed" basis. All onsite emergency centers are located in Unit 1. 4.7.1 Licensee Onsite Emergency Centers 4.7.1.1 Emergency Control Center The designated location for the Emergency Control Center is in the Unit 1 Control Room and the adjacent Sh i f t Supervisor's Office. These areas are designed to protect personnel from radiation hazards and natural phenomenon. Command and control of 111 site-related emergency ef fort s originate from this center. 4.7.1.2 Technical Support Center The Technical Support Center is located in close proximity to the TMI Unit 1 Control Room. It is comprised of two areas: 4-106 Am. 4 1457 075

namely the area just below the Control Room on the 338'6" elevation (Remote Shutdown Station) and the computer room /obser-vation booth adjacent to the Control Room. The area below the Control Room contains the necessary controls to bring the reactor to a safe shutdown. A display unit and patch board provides access to over 100 key plant parameters that are es sential for assessing accident conditions. Plant drawings and supporting information are readily available to occupants of this area. Technical personnel with computer related respon-sibilities will be located in the computer room / observation booth area. Both areas are designed to the same habitability standards as the control room. The purpose of the Technical Support Center is to provide an area outside of the Control Room that can accommodate personnel act ing in support of the command and control functions by furnishing more indepth diagnostic and corrective engineering assistance. 4.7.1.3 Operations support Center The primary location for the Operations Support Center is in the area of the Health Physics Access Control Point on the 306' elevation of the Control Building. The secondary location for this Center is in the Instrument Laboratory and Instrument Supervisor's Of fice which are both adjacent to the Control Room. The purpose of establishing an Operat ional Support Center is to provide an area for shif t personnel to muster for subsequent 4-107 Am. 4 1457 076

assignment to dut ies in support of emergency operat ions. 4.7.2 Licensee Offsite Emereencv Centers 4.7.2.1 Offsite Emergency Supeort Center The TMI Observation Center fronting on Highway 441 east of the TMI site will house the primary Offsite Emergency Support Center. This facility is normally manned as a public relations center and is a well built pe rmanent structure with adjacent parking areas. Sufficient area for helicopter landing is available in close proximity to the TMI Observation Center. The TMI Observation Center will house the key technical groups of the offsite e=ergency organization. Depending upon the logistics requirement , certain non-technical functions will be located either at the Crawford Station or the 500kV Substat ion, south of the TMI Observation Center. (See Figure 13) 4.7.2.2 Backuo offsite Emergency Support Center The Crawford Station which is located approximately three miles north of the TMI site, serves as the backup location for the Offsite Emergency Support Center. 4.7.3 County, State, and Federal Energencv Centers 4.7.3.1 County Emergency Centers Potential TMI-related emergencies could impact the people in Dauphin, Lancaster, York, Cumberland, and Lebanon count ies. 4-108 ^*' 4 1457 077

Each of these jurisdict ions have Emergency Operat ions Centers (EOC's) that meet or exceed the minimum federal criteria for suf fic ient space, communications, warning systems, self suf-ficiency in supplies and accomodations, and a protection factor of 40 or more. All counties maintain at least two full t ime employees to coordinate emergency planning and execut ion. County EOC's are located in the following places:

1. Dauphin County EOC Dauphin County Court House Front & Market Streets HarrisPurg, Pennsylvania 17101

2. L. icaster County EOC Lancaster County Court House 50 North Duke Street Lancaster, Pennsylvania 17602

3. York County EOC York County Court House 28 East Market Street York, Pennsylvania 17401

4. Cumberland County EOC Cumberland County Court House High and Hanover Streets Carlisle, Pennsylvania 17013

5. Lebanon County EOC Lebanon County Municipal Building 400 South 8th Street Lebanon, Pennsylvania 17042 4-109 Am. 4 1457 078

4.7.3.2 State Emergency Center The State Emergency Operations Center (PEMA headquarters) is located in the Transportation Building in Harrisburg. It contains provisione, accomodations, and its own water supply to support State operations people. A reliable communications system ties all area and county emergency operations centers into the State center. During an emergency, representat ives trom all State agencies assemble in the State EOC to manage the response efforts. The State EOC also has the capability to transmit over the emergency broadcasting system. 4.7.3.3 Federal Emergency Center The Federal Emergency Center will be located at the Capital City Airport. This center will have communications ties with the State Bureau of Radiation Protection. Both the Department of Energy and the Environmental Protection Agency radiation assistance terms will operate out of this locat ion. 4.7.4 Media Center The Media Center will be located in a trailer parked adjacent to the TMI Observation Center. Equipment and facilities are designed to support timely communications and information dissemination on plant conditions and emergency ope: 2t ions . Large press conferences will be held in the Hershey Motor Lodge Convent ion Center. Additior.al information on the media center is provided in the Emergency Communications Plan which is attached as Appendix B. 4-110 1457 079

4.7.5 COMMUNICATIONS SYSTEMS The TMI coc:municat ions system is designed to ensure the reliable, timely flow of information and action directives between all parties having jurisdiction and a role to pla; in the mitigation of emergencies at the TMI Nuclear Station. Reliability is provided via (1) extensive redundancy, (2) alternative communications methods, (3) dedicated communication equip-ment to preclude delays due to system swamping, and (4) rout ine use of many of the systems which lowers the probability of undetected system failures. Timeliness of information flow is achieved by (1) prompt notification (2) predefined lines of communication, (3) predefined emergency action levels, and (4) predefined levels of authority and responsibility. Figures 19 and 20 are block diagrams shvving the fundamental communica-tions paths between all parties involved in accident mitigation. The Unit 1 Control Room is the primary source of plant information. Infor-mation originating in the Control Room can be classified into two major catagories, namely, operat ional data and radiological data. The TMI communications network is formulated around this basic concept and is designed to channel informat ion directly tc the key parties having closely related functions thus eliminating errors of ten associated with second hand information. By providing well defined and dedicated com-munication links, better accident management from physically separate control and support centers can ce achieved. 4.7.5.1 Operational Line The " Operational 1.ine" is a special telephone located in the Shift Supervisor's Of fice that has dedicated telephone lines connect ing 4-111 Am. 4 1457 080

the TMI Unit 1 Control Rr - with (1) the NRC office in tethesda, Maryland; (2) the Babcock & Wilcox Company in Lynchburg, Virginia; (3) the TMI Technical Support Center, and (4) the Offsite Ewergency Support Center. The "Operat ional Line" permits an unimpeded flow of plant parameters, system status data, reactor core conditions, and any other information needed by the involved parties to resolve problems in accident mi t i g.i t ion . Conference capability also pe_mits discussions involving all part ies and t imely decisions on subsequent plant operations. The Communicator is responsible for establishing and controlling this special line of communications. 4.7.5.2 Radiological Line The " Radiological Line" is a special telephone located in the Control Room which has dedicated telephone lines connecting it with (1) the State Bureau of Radiation Protection, (2) the NRC Region I Office in King of Prussia, Pennsylvania; and (3) the Offsite Emergency Support Center. The " Radiological Line" permits tha transtission of plant conditions, dose project ions, of fsite monitoring result s, and liquid ef fluent release data to all parties invol7ed with accident assessment. The Radiological Assessment Coordinator, located in the Control Room area, is responsible for establishing and controlling this special line of communication. Conference capability also permits discus-sions involving all parties. 4-112 Am. 4 1457 081

4.7.5.3 NRC Hot-Line The NRC hot-line (red phone) is a dedicated telephone syacem that connects TMI with NRC headquarters in Bethesda, Maryland. The purpose of this line is to provide reliable communications with the NRC. NRC hot-line phones are located in the Control Room, Sh if t Supervisor's Office, Operational Support Center, Technical Support Center and Offsite Emergency Support Center. 4.7.5.4 NRC SS4 Line The NRC SS4 line (black phone) provides a second dedicated line which connects TMI with the NRC in either Bethesda, Maryland or the Region I Office in King of Prussia, Pennsylvania. The purpose of this line is to provide reliable communications with the NRC. NRC SS4 line phones are located in the Shif t Supervisor's Office and the Offsite Emergency Support Center. 4.7.5.5 . National Warning System (NAWAS) Line A National Warning System (NAWAS) telephone connecting the Control Room directly to the Pennsylvania Emergency Management Agency operations center in Harrisburg, Pennsylvania ensures a reliable means for prompt notification of an emergency and, as appropriate, the subsequent exchange of information. 4.7.5.6 Pennsv1vania Bell System TMI is served by the Pennsylvania Bell System for normal tele-phone service. 1r.is system is expected to funct ion during emergencies as it does during normal plant operations. Met-Ed leases tie lines that ensure reliable acces to other locations. 4-113 Am. 4

                                                                           ,1 a

These tie lines preclude local phone activity from swamping calls into and out of the TMI Nuclear Station. 4.7.5.7 Microwave Svstem TMI maintains telephone communication with the entire GPU system via a Company owned microwave transmission system. Access to this mode of transmission is made via the plant telephone system. 4.7.5.8 Radio Communications Radio communication equipment used in normal plant operations will be used in an emergency to communicate with mobile units and to provide backup to the telephone system if necessary. Radio capability includes the following: ,

1. TMI Utility Frequency

2. TMI Security Frequency

3. Met-Ed System (Lebanon Frequency)

All three systems are used daily and are highly reliable. Remote control stations locatad in the Control Room and the Operations Support Center permit transmission on the TMI utility frequency. The Control Room unit is equipped with a device that can trigger a signal to be transmitted that will result in the Dauphin County Civil Defense scanr.er locking onto the TMI ut ili t:- frequency. Radio communicstions can then proceed with TMI transmitting on the utility frequency and receiving on a scanner which monitors the Civil Defense frequency. 4-114 Am. 4 1457 083

4.7.5.9 Inter-Control Room Hot-Line A direct line between the Unit 1 Control Room and Unit 2 Control Room provides a reliable path for communications between Control Rooms. 4.7.5.10 Emergency Director's Hot-Line A dedicated phone line has been established between the Contrcl Room of Unit I to the Of fsite Emergcncy Support Center. This phone i, located in the Shift Supervisor's office. 4.7.5.11 Alarms Audible alarms are a quick and effective means of co municating emergency warnings on the Thi site. Alarms currently installed at TMI include the:

1. Radiation Emergency Alarm

2. Fire Alarm

3. Reactor Buil.iing Evacuation Alarm Each alarm provides a distinctive sound that all TMI Nuclea-Station personnel and contractors are trained to recognize.

4.7.5.12 Plant Paging System The plant paging system has two subsystems, namely, the Intra-Plant Communications Subsystem and the Redundant Communication Subsystem. The Intra-Plant Communications Subsystem links together permanent plant structures through a network of phone stations and speakers. This subsystem provides four channels of coc:munication, one page channel and three party channels. The Redundant Co==unication Subsystem serves as a separate, indepen-4-115 Am. 4 1457 084

dent system in the event of failure of the Intra-Plant Communi-cations Subsystem. Phone stations and speakers of this subsystem are 1ccated in vital plant ares.s.

4. 7.5.13 Maintenance and Instrumentation Phone System The Maintenance and Instrumentation Phone System consists of three essentially independent circuits; the Nuclear Subsystem, the Turbine Subsystem, and the Fuel Handling Subsystem. These circuits are for use between two or more locations during operations when direct communications between operators and/or maintenance personnel is tequired. Handsets and headsets are provided. The system is operable when headsets and/or handsets are plugged into the various stations of the three subsystems.

4.7.6 Assessment facilities 4.7.6.1 Onsite Systems and Ecuipment

1. Radiation Monitoring System The onsite Radiation Monitoring System contributes to personnel protection, equipment monitoring, regulatory compitanca, data gathering, and accident assessment by measuring and recording radiation levels and concentrat ions of radioact ive material at selected locations within the plant. The Radiation Monitoring System detects, alarms, and initiates, if required, emergency act ions when radiat ion levels or radionuclide concentrations exceed predetermined levels. To perform these funct ions , area, liquid, and atmospheric monitoring subsystems are required.

4-116 Am. 4 1457 085

The data from theme subsystems are displayed by readouts in the Control Room and are recceded by strip chart recorders which are also located in the Control Room. A summary description of individual radiation monitor channels, which are generally described below, is provided in Table 9. In general the radiation monitoring equipment is designed in accordance with the following: o Each monitoring station has adjustable, high trip, low trip, and power supply failure alarms. o Solid-state circuitry is used except for primary detectors. o All AC operated radiation monitoring equipment , ex';e pt for the pump assemblies , is provided with power from the battery-backed, inverter-fed vital power supply busses. The sampling station pumps are fed from the engineered safeguards power systems. o Each radiation monitor is capable of being ch(cked periodically with solenoid actuated check sources. o A pulse generator or current source is used for electrically checking each monitor or subsystem. Electrical input test s measure the funct ional oper-ation of the monitoring system from the detector output thru the readout devices. o The modules are designed so that an alarm and/or indicat ion is init iat ed when f ailure occurs any-where in the channel. 4-117 Am. 4 1457 086

a. Area Radiat ion Monitoring The area radiation monitoring subsystem is comprised of channels which utilize an ionization chamber type detector housed in a weatherproof container and equipped with a remote controlled check source.

The local alarm and readout for each of these channels is separate from the detector and is also housed in a weatherproof container. One channel is provided to monitor the high level of radiation that would be characteristic of the post-accident atmosphere in the containment . The detector is densensitized by a lead shield. The Control Room readout modules are located in the radiation monitoring panel in the control room.

b. Atmospheric Radiation Monitoring The atmospheric radiation monitoring subsystem is comprised of monitors of the fixed and movable type. Each fixed atmospheric monitor is comprised of a particulate measuring channel, iodine measuring channel, and a gaseous measuring channel. The air sample that passes through each of these channels is obtained by means of an isohinetic sampler and a pump assembly. Samples are obtained by means of a sampling head placed in a ventilation duct . The sampling heads are equipped with one or more isokinetic 4-118 Am. 4 1457 087

nozzles such that, with a sample ' low of I cfm, the velocity of the air sampled is the same as the velocity of the air in the duct. Movable atmospheric monitors on carts are typically used in

  - ~ - - --

the spent fuel handling area during refueling operations and in the radiochemical laboratory during laboratory sample preparation operations. These monitors are also supplemented oy various portable radiation monitoring equipment. Each monitor contains three channels for particu-late, iodine, and gaseous monitoring, respect ively. Visual high-radiation alert / low-level alarms are provided for each channel at the local monitor sampling station for all atmospheric monitors. Each channel shares an audible alarm at the local monitor sampling station.

c. Liquid Radiation Monitoring The liquid radiation monitoring subsystem consists of moritors each of which consists of a sampler, scintillation detector, and Control Room ratemeter module. The monitors indicate in the Control Room on the individual ratemeter modules and two common recorders. The Primary Coolant Letdown monitor also contains a high range channel consisting of a GM detector that monitors the same sample but uses a separate Control Room ratemeter module.

4-119 Am. 4 1457 088

2. Fire Detection Devices Fire protection at TMI is provided by (1) the Fire Service Water System, (2) the Halogenated Fire Suppression System, (3) the CO2 Fire Extinguishing System, and (4) the Dry Chemical System.

The Fire Service Water System is a full-loop, ;= ped system that supplies water for (1) sprinklers, (2) deluge water spray, (3) fire hydrants and (4) hose connections that are located such that they provide fire protection for all major areas of the plant and site. A 100,000 gallon Altitude Tank provides the source of water to maintain system piping full and pressurized. This provides a method of monitoring system piping integrity and provides an initial supply of water for suppressing a fire. In the event a fire occurs, and either an automatic or manual system is initiated, the Fire Service Water System piping pressure will decrease which will cause, seqs ucially, one electric and two diesel fire pumps to start as necessary to meet system flow requirements. The Halogenated Fire Suppression System is provided in the plant ventilating air intake and tunnel to inhibit combus-tion of any fuel / air mixture which might enter the intake structure. Detection of an embryonic explosion releases Halon 1301 gas into the mixture in sufficient quantity to render it incombustible within a fraction of the time 1457 089 4-120 Am. 4

required for the explosion to reach destructive proportions. The CO 2 Fire Extinguishing System provides fire pro-tection for the 338 foot Elevation Relay Room. A supply of low pressure CO 2 is maintained in a refrigerated storage unit. The CO2 is discharged into the Relay room, after a time delay, when a manual pushbutton is depressed or when a thermostat indicates high room temperature.

a. The above systems are tctuated either manually or automatically in response to signals from detectors monitcring conditions in the protected area. The following is a list of detectors used in various systems discussed above:

(1) Thermostats Thermostats are used to monitor the temperature in the protected area. When the temperature in the area exceeds the setpoint of the thermostat, a signal is either sent to an automatic control circuit which will function to suppress the fire or sound an alarm causing an operator to activate a suppression system, or both. (2) Embryonic explosion detector The embryonic explosion detectors consist of a photocell and a pressure detector. The photocell will sense the initial flash of a flame front, and 4-121 Am. 4 1457 090

the pressure detector will sense *he pressure wave as the flame propogates. This detector will react to an explosion before the explosion has time to reach destructive proportions. This signal is sent to the Halogenated Fire Suppression System, . and will alarm in the Centrol Room. (3) Temperature rise detectors Temperature rise detectors wil? monitor the protected area, and send a signal to either an automatic control circuit or an alarm in the Control Room, or both, if the rise in temperature reaches a setpoint. (4) Fusible link Fusible links are used in sprinkler systems to hold the sprinkler heads closed. When the tempera-ture at the sprinkler head rises to the preset tempera ture, the link melts releasing the sprinkler head which then opens. In the case of a wet sprinkler system, water is then discharged through the sprinkler to suppress the fire. In the case of a dry sprinkler system, air is Fled out through the sprinkler head, which depressurizes the sprinkler header and causes a water supply valve to open thereby supplying water to the sprinkler, i 4-122 Am. 4 1457 09i

(5) Flow detectors Flow detectors are used in the sprinkler systems (refer to (4) above), to sense the flow of water through a header supplying a sprinkler system, there-by indicating thst the sprinkler system has been actuated. This flow detector will send a signal to an alarm in the Control Room to alert the operator as to the status of that sprinkler system. (6) Smoke and vapor detectors Ionization type smoke and products of combustion detectors are installed in the Control Building air return ducts. When smoke or product s of combustion are detected, an alarm will sound in the Control Room and the ventilation system will aut7matica11y shutdown. Combustible vapor detectors are used to monitor Auxiliary Building plant ventilation exhaust ducts.

3. Seismic Monitoring Strong motion recording systems at the TMI Nuclear Station ceasure ground mot ion and structural vibrating response caused by an earthquake occurring in the vicinity of the site. A cassette magnetic tape recorder located in the Unit 1 Control Room receives information supplied by the triaxial sensor units which are firmly mounted on the 4-123 Am. 4 1457 092

Reactor Building. One triaxial sensor unit is attached immediately outside of the contcinment wall at the base of the Reactor Building. A second triaxial sensor is situated along the same Reactor Building axis, but is attached to the Reactor Building ring girder. The triaxial sensor units begin to supply seismic data to the magnetic tape recorder after a signal is sent to the sensors by a remote starter unit. A remote starter unit attached to the base of the Reactor Building provides a signal for its systems sensor units when the starter unit detects a ground acceleration greater than a preset threshold level. The remote starter also actuates an annunciator labeled " Threshold Seismic Condition". If the ground acceleration exceeds the horizontal or vertical setpoints, a seismic trigger, also mounted on the base of the Reactor Building, will cause a Control Room annunciator labeled " Operating Basis Earthquake" to actuate. The time history of a ground mot ion and result ing vibrating response can be displayed by usi.ig magnetic tape cassettes containing the recorded data, and the magnetic tape playback system in the Control Room. The magnetic tape playback system produces visual playouts of selected magnetically recorded data. This is accomplished with a strip chart recorder built into the playback system. A visual playout allows quick analysis of the earthquake. The magnet ic tapes are 4-124 Am. 4 1457 093

available also for detailed analysis. Peak reading accelerographs are anchored in the Station to Class I selected items. These accelerographs will produce a permanent record of the peak amplitude of the low frequency accelerations caused by seismic disturbances.

  'Ihis record is in the form of magnetic erasure clips which must be developed, using the magnetic developer kit.

After developing, these clips can be s tamined to verify seismic response which had been determir ed analyt ically.

4. Meteorological Monitors - Onsite The onsite meteorological tower, located at the north end of the island, is 150 feet high. This tower provides redundant instrumentat ion that continuously monitors (1) wind speed and wind direction at the 100 foot level and (2) wind direction range and temperature difference between 150 feet at.d 33 foot levels. Readings are continuously recorded on strip charts in the Control Room, and data from each sensor is satapled by a Sperry-Univac mini-computer under control of sof tware which enables remote access, provides calibration checks and prints system quality control diagnostics.

5. Process Monitors Process monitors measure appropriate parameters that are indicative of the status of various plant systems and the reacter itself. In addition, these parameters are displayed 4-125 Am. 4 1457 094

in the Control Room. Table 11 provides a listing and a description of Control Room Instrumentation that would, as appropriate, be used in performing emergency assessment act ivit ies .

6. Laboratory Facilities The TMI 'Jnit 1 laboratcry facility is equipped to provide the water chemistry and radiochemical analysis support required during normal plant operations. A list of equip-ment maintained in the laboratory is included as Table 13.

4.7.6.2 Facilities and Eauipment for Offsite Monitoring A complete Radiological Environment Monitoring Program (REMP) for the Three Mile Island Nuclear Station has been established by Metropolitan Edison Company. Much of the program has been in effect since June 1969. The objectives of the REMP are:

1. To fulfill the obligations of the radiological environ-mental surveillance sections af the Environmental Technical Specifications (reference 10.13.2).

2. To determine whether any statistically significant increase occurs in the concentration of radionuclides in critical pathways.

3. To detect any buildup of long-lived radionuclides in the environment.

4 To detect any change in ambient gamma radiat ion levels . 4-126 Am. 4 1457 095

5. To verify that radioactive releases are within allowable limits and that plant operations have no detrimental effects on the health and safety of the public or on the environment.

Samples for the REMP are taken frem the aquatic, atmospheric, and terrestrial environments. Sample types are based on (1) established critical pathways for the transfer of radio-nuclides through the environment to man, an2 (2) experience gained during the preoperational and initial operatior.al phases of the REMP. Sampling locations were determined from site meteorology, Susquehanna River hydrology, local demography, and land uses. Sampling locations are divided into two classes; indicator and control. Indicator stations are those which are expected to monitor plant effects, if any exist; control samples are collected at locations which are believed to be unaf fected by plant operations. Fluctuations in the levels of radionuclides and direct radiat ion at indicator stations are evaluated with respect to analogous fluctuations at control stat ions. Indicator station data is also evaluated relative to background charac-teristics established prior to station operation. This media sampled are: o AI = Air Iodine o FPL = Green Leafy Vegetables o = AP Air Particulates o ID = Immersion Dose (TLD) 4-127 . Am. 4 1457 096

o AQF = Fish o M = Milk o AQP = Aquatic Plants o RW = Precipitation o AQS = Sediment o SW = Surface Water o E = Soil o V = Fodder .;rops o FPF = Fruit o MG = Milk (Goats) Table 10 summarizes the type of samples collected, collection methods used, and the analysis that is performed as a part of the REMP. The thermoluminescent dosimeter (TLD) program used by Met-Ed meet s the requirement s set forth in the TMI Environmental Technical Specifications as recommended by the Nuclear Regulater Commission. The TLDs are located such that at least two are situated in each 22 1/2 sector surrounding the TMI site, with one on the site boundary and the other 4 to 5 miles out from the site. Other TLDs are located in areas of interest (i.e. near more densely populated areas , etc. ) . During normal operations accurate measures of the environmental dose rate can be made and during an accident situation time incremental and t ime integrated doses can be made. 4.7.7 Protective Facilities and Ecuipment Personnel protective action at TMI is a function of the nature of the hazards, (i.e., preparing for a hurricane is somewhat different from preparing lor radiological hazards). Preplanned responses to the basic 4-128 Am. 4 1457 097

hazards, high wind, flooding, earthquakes, and radiation exposure are an integral part of the TMI Emergency Plan and are detailed in other sections. A fundamental concept in personnel protection is the immediate release and removal of all individuals not essential to the operation, safety, security, and damage control of the plant. Obviously some hazards can occur before any significant protective action can be applied; an earthquake for example. When the situation permits positive action, the appropriate alarms are sounded and all personnel on the site either assume their assigned emergency responsibilities or are assembled at the designated points for accountability prior to release or assignment to an emergency tea . The Process Center (security processing facility) is the assembly point for all personnel outside of the protected area in the north east section of the plant complex. This facility is a pre-engineered metal building with conventional ventilation equipment capable of handling several hundred peopla. t swer facilities for both men and women are providtd in locker rooms. Because ef fluents from these showers drain to the plant sewer system they will be used for personnel decontamination only in an extreme emergency. Minor first aid equipment is available at this assembly point. The Unit 1 Service Building Auditorium is the assembly point for people inside the protected area of TMI Unit 1. This facility will accomodate several hundred people and has an adjacent first aid station. Showers that drain to the plant sewer system are provided for both supervisors and bargaining unit personnel. 4-129 Am. 4 1457 098

ne Met-Ed Unit 1 Warehouse is the assembly point for personnel out-side the protected area in the northwest section of the plant site. h is structure is a pre-engineered metal building with a conventional vent ilation system. It is normally manned at all times . Respirators, protective clothing, and most other protective equipment for the plant are stored in this warehouse. If required, personnel assembled at this point could be issued protective equipment from stored supplies. Warehouse No. 2, located west of the Unit 2 Turbine Building is the assembly area for all personnel outside of the protected area and south of the fence separating Unit I and 2. The ground level of the Unit 2 Turbine Building is the assembly area for all personnel within the Unit 2 protected area. Protective facilities include the TMI Unit 1 Control Room. This area is located in seismically rated structures and have adequate shielding to permit safe occupation for extended periods of time without exceeding an exposure limit of 3 Rem. The Control Room ventilation system has redundant fans and chillers and is provided with radiation and smoke detectors with appropriate alarms and interlocks. Provisions have been made for the Control Room air to be recirculated through high ef ficiency particulate (HEPA) and activiated charcoal filters. Fresh air is drawn through an underground ventilation tunnel which has been provided with protection against combustible vapors, incipient explosions or fires. The tunnel is seismic class I and is also designed for the hypothetical aircraft incident. Scott air packs and respirators are located in the Control Room to 4-13 0 Am. 4 1457 099

pe rmit continued occupancy if ventilation systems fail. An extensive medical aid facility is located below the Control Room in the Health Physics Lab. 4.7.8 First Aid and Medical Facilities First aid facilities at TMI are designed to support a wide range of immediate care ranging from simple first aid up to and including pro-cedures requiring a physician. The most readily available first aid is provided by small kits placed throughout the plant. These kits contain 16 items (units) typically needed to care for minor injuries. The next level of first aid equipment is found at twelve first aid stations located in the following areas: o Unit 1 Circulating Pump House o Training Department Hallway o Unit 1 Health Physics area hallway o Unit 1 River Water Pump House o Industrial Waste Building o Unit 2 Turbine Building near the roll-up door o Unit 2 Control Room o Unit 2 Reactor Building entrance o Unit 2 Auxiliary Building, elevation 305 feet, by the elevator o Unit 2 Circulating Water Pump House o Warehouse No. I north end at the receiving dock o Hallway in the Unit 1 Control Building These stations are equipped with sufficient quantities of the following 4-131 Am. 4 1457 100

supplies to serve the expected needs of 100 employees: o Stretcher o Inflatable Arm Splint o Inflatable Leg Splint o Utility Blanket o Large Plastic Sheet o 100 Employee Si:e First Aid Kits (contents lised in Taole 14) Four specific areas in the TMI Nuclear Station have been set aside as first aid f acilities. In a medical emergency the particular facility used would depend upon (1) its proximity to the injured party, (2) the nature of the injury, and (3) the extent of the injury. Unmanned rooms are located in Unit I at the 305' elevation in the east hallway in the Service Building and in Unit 2 on the 305' elevation in the north hallway in the Control B.ilding. These rooms are usually unlocked and are typically out fitted to support the aid that can be administered by personnel qualified to the multi-media level of Red Cross training. Typical First aid supplies found in these rooms are included in Table 15. A first aid facility is located in a trailer on the west side of the main roadway adjacent to the Unit 1 Air Intake. A State registered emergency medical technician mars this facility during normal working hours. Other medical equipment is available at this facility as listed in Table 16. A special medical facility is located in the Unit I health physics area at the 305' elevation to permit the treatment of contaminated / injured personnel. This facility is normally locked with keys maintained at the Access Control Point and in the Control Room. Equipment inventory 4-132 Am. 4 1457 101

is planned to support professional medical treatment and includes the items listed in Table 17. 4.7.9 Damage Control Equipment The TMI Nuclear Ecation is extensively equipped to conduct preventive maintenance and repairs on mechanical, structural, electrical, and instrumentation and controls equipment found in the plant. Operational policy provides a minimum maintenance crew assigned to the onsite shift organizations at all times. Each individual assigned to the maintenance crew is qualified and, when required, certified to perform the t asks associated with his craft in the working environment of a nuclear plant. In addition to the equipment and materials required for normal mainten-ance, other items are available to handle extraordinary maintence jobs that might arise in datnage control. Thes; items are listed in Table 18. Select ion of damage control equipment inventory is based upon (a) miti-gating the consequences of flooding, (b) personnel rescue, (c) checking the uncontrolled flow of fluids from process systems, and (d) elimination of electrical hazards. 4-133 Am. 4 1457 102

4.8 MAINTAINING EMERGENCY PREPAREDNESS Metropolitan Edison Company will maintain, as two separate documents, this Emergency Plan and the Emergency Plan Implementing Document. It is intended that this Plan, although considered as ;. art of the Three Mile Island Nuclear Station Unit . Final Safety Analysis Report (FSAR), will be me,intained as a separate document as suggested by the guidance provided in reference 10.4. Ef fort s will be made to assure continuous emergency preparedness and operational readiness among Met-Ed personnel and the offsite response agencies and organizations. The Met-Ed Vice President of Nuclear Operations has been assigned overall responsibility for emergency planning related to the TMI Nucisar Station. This responsibility includes not cnly the TMI Emergency Plan and Implementing Document but also includes its interrelationship with State, federal, and county plans; agreemen letters; corporate policy and plans; and other related plans, programs, and procedures. To assist the Vice President of Nuclear Operation in meeting his assigned responsibilities, an Emergency Planning Coordinator has been designated. The specific responsibilities delegated to the individual assigned as the Emerge.ncy Planning Coordinator are described in the following subsections and in particular, subsection 8.1.3. 4.8.1 Organizational Preparedness 4.8.1.1 Training All personnel at the Three Mile Island Nuclear Station will take part in a formal training program under the direction of the Manager, Training. In general, this training program provides for the indoctrination of Met-Ed/GPU employees and contractors 4-134 Ar, 4 1457 103

in addition to providing specialized training for licensed operators, health physics / radiation protection personnel, and personnel assigned specific responsibilities in the emergency organization. "Ihe Manager of Training is responsible for ensuring that personnel in each department receive the appropriate training. He may delegate specialty training responsibilities to personnel qualified to perform such training. The training program for the TMI Nuclear Station with regards to the TMI Emergency Plan will include the following:

1. All Three Mile Island staff personnel, except personnel in the Operations Department, are required to attend the General Employee Training Program at least once per calendar year. In addition, the prompt indoctrination of new employees and i.ontractor personnel is provided for in the Health Physics Training Program which they are required to attend prior to receiving the privileFe of unescorted access onsite. With regards to emergency planning, the objectives of these programs are to:

a. Familiarize personnel with the scope, applicability, and implementation of the TMI Emergency Plan and Implementing Document.

b. Teach the general duties and responsibilities assigned to all TMI personnel.

c. Keep personnel informed of any changes in the TMI Emergency Plan and/or the Implementing Document.

d. Maintain a high degree of preparedness at all levels of the TMI Nuclear Station organization.

4-135 Am. 4 1457 104

To meet these objectives, each IMI Nuclear Station empicyee and unescourted contractor personnel will receive, as a minimum, the following instruction:

a. Orientation in the content of the TMI Er.:ergency Plan and the Implementing Document.

b. Orientation in the implementation and operation of the TMI Emergency Plan, including the assignment of duties and responsibilities, location of emergency centers and assembly facilities, and the location of e=ergency equip-ment and supplies, where applicable.

c. Orientation in individual employee responsibilities with regard to the use of emergency facilities and equipment, familiarization with stat ion alarms and personnel response, and the use of general station communications systems.

d. Orientation in instructions and requirements associated with personnel accountability, evacuation, and exposure criteria.

e. Orientation in radiation protection with emphasis on the principles and use of protective clothing, equipment, and personnel dosimetry.

2. Personnel assigned to the Met-Ed emergency organization with specific Emergency Plan duties and responsibilities shall receive specialized training for their respective assignments. Table 12 delineates which personnel will receive specialized training, 4- 13 6 A=. 4 1457 105

the type of training, and the minimum required frequency of such training.

3. The Pennsylvania Emergency Management Agency (PEMA) develops, conducts, coordinates, and promotes a training program throughout the State and assists the counties in developing training policy for disaster operational readiness. The county and local Civil Defense Directors are responsible for planning and conducting disaster preparedness traintag of respective emergency response personnel.

Met-Ed will work closely with PEMA and the county and local Civil Defense Directors in coordinating training programs. In addition, orientation and training of State and county agencies and personnel involved in TMI emergency planning ef forts will be made available by Met-Ed.

a. The civil defense organizations listed below will be invited, on at least an annual basis, to participate in a training program at the TMI Nuclear Station. The program will relate the importance of ef fective planning for emergency situations and interfaces between the " licensee's" emergency organizations and the of fsite (i.e. State, county, and federal) emergency organizations. The program will also include a review of the TMI Emergency Plan and Implementing Document with particular emphasis given to the classification of emergencies; reporting requirements; assessment, protective, and corrective actions; and communications networks.

4-137 1457 106

and communications networks. (1) Pennsylvania Emergency Management Agency (2) Dauphin County Civil Defense (3) York County Civil Defense (4) Lancaster County Civil Defense (5) Cumberland County Civil Defense (6) Lebanon County Civil Defense

b. At least annaally, the State Bureau of Radiation Protection will be invited to part icipate in a training program at the TMI Nuclear Station. The program will, as does the program for the civil defense organizations, relate the importance of ef fective planning for emergency situations, the inter-face between the " licensee's" emergency organizations and the of fsite (i.e. State, cour*y, and tderal) emergency organizations. The program will include a review of the TMI Emergency Dlan and Implementing Document with particular emphasis given to the classification of emergencies; reporting requirements; assessment, protective, and cor-rect ive act ions ; and communications networks. In addition, specific training on dose calculations / projections, pro-tect ive action guides, and reportable information will also be provided.

c. The State Police will, on at least an annual basis, be invited to participate in a training program that will include a review of the applicable parts of the TMI 4-138 Am. 4 1457 107

Emergency Plan and Implementing Document with emphasis on the classification of emergencies, communications, and specific areas of responsibility.

4. Met-Ed will also provide orientation and training to local services support organizations as specified in respective letters of agreement and 4 required to ensure a high state of emergency preparedness and response capability between these organizations aad the TMI Nucle:r Statien emergency organization.

The local services support organizatio , and personnel who may provide onsite emergency assistance will be encouraged to become familiar with the ~'MI Nuclear Station (including the physical plant layout) and key station personnel, and will be invited to attend Emergency Plan orientation and training courses conducted by or for Met-Ed It is anticipated that such training will be pro-vided on at least an annual basis and will be made available to the appropriate personnel of the following organizations and certain specified individuals.

a. Middletown Police Department The Middletown Police Department will be invited to part icipate in a training program that will include a review of the applicable parts of the TMI Emergency Plan and Implementing Document with emphasis on the classification of emergencies , com:nunicat ions, and specific areas of emergencies, communications, and specific areas of respon-sibility.

4- 13 9 Am. 4 1457 108

b. Fire Companies The local fire and rescue companies will be invited to par-ticipate in a training program that will, as a minimum, include the following topics:

(1) Interface with the Site Security Force during emergencies. (2) Basic health physics indoctrination and training. (3) TMI Nuclear Station facility layout. (4) Onsite fire protection system equipment (permanent and portable) . (5) Differences between onsite firefighting equinenent and fire company supplied equipment. (6) Communications systems. (7) Review of applicable parts of the TMI Emergency Plan and Implementing Document. (8) The onsite emergency organization with specific emphasis on the interface between the TMI Fire Brigade and Fire Company personnel.

c. Hershey Medical Center, Local Physicians, and Fire Company (listed above) Ambulance Services The local medical support organizations and personnel will also be invited to participate in a training program that will, as a minimum and as applicable, incluse the following topics:

4-140 Am. 4 1457 109

(1) Interface with tha Site Security Force during emergencies. (2) Basic health physics indoccs ination and training (3) TMI Nuclear Station facility layout (4) Onsite medical treatment facilities, equipment, and supplies (5) Communications systems (6) The onsite emergency organization with specific emphasis on the interface between the TMI First Aid and Rescue Team (s), the local medical support personnel, and Radiation Management Corporation. (7) Radiological aspects of emergency medical treatment. (8) TMI Nuclear Station procedures for decontaminat ion. (9) Hershey Medical Center radiation emergency procedures. 4.8.1.2 Drills and Exercises Periodic drills and oxercises will be conducted in order to test the state of emergency preparedness. The prime objective of this form of training is to verify the emergency preparedness of all participating personnel, organizations, and agencies. Each drill or exercise will be conducted to: (1) ensure that the participant s are familiar with their respective duties and responsibilities, (2) verify the adequacy of :he TMI Emergency Plan and the methods used in the Emergency Plan Implementing Procedures, (3) test communications networks and systems, (4) check the availability of 2mergency supplies and equipment, and (5) verify the oper-ability of emergency equipment. 4-141 Am. 4

                                                                  \451 \\D

The Emergency Planning Coordinator will be responsible for the planning, scheduling, and coordinating of all emergency planning related drills and exercises. The Manager of Training will assist the Emergency Planning Coordinator in carrying out these *sponsibilities, however, all drills and exercises are subject to the approval of the Unit 1 Superintendent. In addition the Vice President of Fuelear Operations shall approve the annual Radiation Emergency Drill. When a major drill or exercise is to be conducted, the Emergency Planning Coordinator will:

1. Assign p;rsonnel to prepare a scenario.

2. Coordinate ef forts with other participating emergency personnel, organizations, and agencies.

3. Obtain the approval of the Unit 1 Superintendent, and (for the annual Radiation Emergency Drill) the Vice President of Nuclear Operations.

4. Schedule a date for drill execution and assign observers.

5. Critique the results of the drill.

6. Assign personnel to correct any deficiencies.

7. Ensure that deficiencies are corrected.

8. Prepare and submit documentation to the Training Department for recordkeeping.

Scheduled drills and exerciees will be held involving appro-priate of fsite as well as onsite emergency personnel, organiza-tions, and agencies. These drills and excercises will be Am. 4 4-142

                                                              ,   .1

conducted, simulating as closely as possible actual emergency conditions and c.ay be scheduled such that one or more drills or excercises can be conducted simultaneously. Drill scenarios can and will be prepared that involve participation of several emergency teams and all or specific parts of the onsite and offsite emergency organizations including varying degrees of participation of State, county, and federal agencies and organizctions and local services support personnel and organizations. The Emergency Planning Coordinator will normally notify the offsite emergency response organizations and agencies at least thirty days in advance of the the scheduled date of the drill or exercise. Recommendations for revisicas to the TMI Euiergency Plan and/or the Implementing Document and/or the upgrading of emergency equipment and supplies as a result of a drill or exercise shall be forwarded to the Emergency Planning Coordinator by observers or participants. The Emergency Planning Coordinator will submit such recommendations to the Unit 1 Superintendent for review. Recommended changes that are approved by the Unit 1 Supe rint endent shall be incorporated into the Emergency Planning Program under the direction of the Emergency Planning Coorinator Records will be maintained on each drill listed below. Major drills and exercises will be conducted as described below:

1. Medical Emergency Drill

a. At least one drill per calender year shall be conducted. The drill will involve the participation of some, if not all, of the local medical support personnel 4-143 Am. 4 1457 112

and organizations (e.g., physician, ambulance service, hospital, etc.), and will involve cases of radiation overexposure and/or contaminated personnel a.d/or con-taminated/ injured personnel.

2. Fire Emergency Drill

a. At least one drill per calendar quarter shall be conducted.

b. At least one diill in the calendar year shall involve the participation of at least one, if not all of the local fire departments.

3. Repair and Damage Control Drill

a. At least one drill per calendar year shall be conducted t' exercise the Emergency Repair Team.

4. Communications Links Teat

a. At least once per calendar quarter, the communication links used for notification (i.e. TMI Control Room to Dauphin County Civil Defense, Pennsylvania Emergency Management Agency, the NRC, the Emergency Director and Emergency Support Director, etc.) shall be tested.

5. Radiation Emergency Drill

a. A major drill appropriate to a Site or General Emergency shall be conducted at least once per calendar year.

b. Conduct of the drill shall provide.for the coordin-ation with and participation of of fsite emergency response personnel, organizations, and agencies 4-144 1457 113

including those of State and county governments. (1) About once every five years a joint exercise appropriate to a Site or General Emergency that involves federal, State, and coanty emer-gency response personnel, organizations, and agencies will be conducted. The scope of the exercise will test as much of the emergency plans (i.e., Met-Ed/TMI, State, and counties) as is reasonably achievable. The degree of public participation in this exercise shall be determined by the appropria'e State agencies. 4.8.1.3 Emergency Planning Coordinator A member of the TMI Nuclear Station staf f will be designated as the Emergency Planning Coordinator. His responsibilities shall include , but not necessarily be limited to:

1. Ensuring the coordination of the TMI Emergency Plan with the:

a. State plans (reference 10.22 and 10.23)

b. County plans (reference 10.24 through 10.28)

c. TMI Security Plan

d. het-Ed Emergency Communications Plan

2. Ensuring that the information, data, and procedures detai.ed in ti.e Emergency Plan Implementing Document are consistent with the evidence providet' in the TMI Emergency Plan.

3. Ensuring that the Emergency Plan Implementing Procedures are coordinated and interface properly with other procedures 4-145 Am. 4 1457 114

3. Ensuring that the Emergency Plan Implementing Procedures are coordinated and interface properly with other procedures (e.g. Administrative Procedures, Security Procedures, Health Physics Procedures, and Training Procedures, etc.).

4. Assisting the Manager of Training in coordinating and/or providing emergency planning related specialty training.

5. Coordinating emergency planning related drills and exercises as described in subsect ion 8.1.2 above.

6. Coordinating the review and updating of the TMI Emergency Plan and Implementing Document as described in Section 8.2 below.

7. Ensuring the maintenance and inventory of emergency equip-ment and supplies as described in Section 8.3 below.

8. Maintaining himself current with respect to changes in federal regulations and guidance that impact emergency plannin activities.

4.8.2 Review and Updating of the Emergency Plan and Imnlementing Document The TMI Emergency Plan, including appended letters of a5reement and plans of of fsite organizations and agencies, will be reviewed and up-dated on at least an annual basis. The TMI Generation Group Technical Support Staff is responsible for auditing, at least once every two years, the TMI Emergency Plan to verify compliance with the TMI Operational Quality Assurance Plan, the Fire Protection Program Plan, internal rules and procedures, federal regulations, and im. 4 4-146 1457 115

operating license provisions. In addition, the Emergency Planning Coordinator will by virtue of his involvement with the TMI Emergency Planning Program, provide an ongoing review. Personnel performing reviews or audits of the TMI Emergency Plan and/or Implementing Document shall take into account Corporate policy, State policy and plans, county plans, and the various agreements and understanding with federal, State, county, and local support agencies and organizations. As previously mentioned, the TMI Emergency Plan is considered a part of the TMI Nuclear Station Unit 1 FSAR. As such, revisions to the Plan shall be administrative 1y controlled by utilizing existing methods which are used in making ammendments to the FSAR. The TMI Emergency Plan Implementing Document will be incorporated into the TMI Nuclear Station procedures program. As such, the Imple-ment ing Document will be prepared, reviewed, approved, controlled, distributed, and revised in accordance with TMI Nuclear Station Admini-strative Procedures. Document holders (e.g. Met-Ed; State, county, and federal agencies; etc.) will receive revisions to the Emergency Plan Implementing Document in a controlled manner as they are issued. In addition, Part I of the Implementing Document will provide guidance to document holders on how to make comments and recommendations concerning the Emergency Planning Program to Metropolitan Edison Company. The Emergency Planning Coordinator is responsible for cc,ardinating the periodic reviews and audits of the TMI Emergency Plan and Implementing Document. In addition, the Emergency Planning Coordinator shall; 4-147 Am. 4 1457 116

through letters, meetings, seminars, or other means available; ensure that all elements of the total emergency organization (e.g. Met-Ed, State, federal, county, etc.) are informed of the TMI Emergency Plan and a=mendments thereto; and the Implementing Document and revisions thereto. Results of each annual review and update will be reported to the Vice President of Nuclear Operations. 4.8.3 Maintenance and Inventory of Emergency Ecuipment and Supolies The Emergency Planning Coordinator is responsible for planning and scheduling the quarterly inventory and inspection of designated emergency equipment and supplies. He will assign personnel to perform these activities. Designated emergency equipment and supplies and their s*orage locations will be listed in the Emergency Plan Implementing Document. Such equip-ment and supplies will be maintained in accordance with approved TMI Nuclear Station procedures. Equipment, supplies, and parts having shelf-lives will be checked and replaced as necessary. Any deficiences found during the inventory and inspection will be either cleared immediately or documented for corrcetive action. A report of each inventory and inspect ion, including documented deficiencies, will be prepared and submitted to the Emergency Planning Coordinator. The Emergency Planning Coordinator will assign personnel responsible for correcting deficiencies and shall ensure that identified deficiencies are corrected in a reasonable period of time. 4-148 Am. 4 1457 117

4.9 RECOVERY The Emergency Director and E.nergency Support Director have the joint responsibility for determining cnd declaring sten an emergency situ-ation is stable and has entered the recovery phase. They will evaluate the status of the emergency by observing monitoring instrumentation and reviewing all current and pertinent data available from emergency response and/or monitoring teams. They shall consider the emergency under control and in the recovery phase only when the following general guidelines are met: o Radiation levels in all in plant areas are stable or are cecreasing with time. o Releases of radioactive caterials to the environment from the plant are under control or have ceased. o Any fire, flooding, or similar emergency conditions are controlled or have ceased. Although planning for recovery will vary according to the specific nature of the emergency situation, a long-term recovery organization that is general in nature has been defined. This organizu ton is depicted in Figure 14 and is described in subsectio .5 of this Plan. During recovery operations, the radiation exposure limits of 10 CFR 20 shall apply. Compliance with thcse limits shall be the responsibility of the Recovery Operations Manager via the applicable Health Physics organization at the time of recovery. Am. 4 4-149

                                                                         \457 \\8

At the time of declaring that an emergency har entered the recovery phase, the Emergency Director shall be responsible for providing noti-fication to all applicable agencies (e.g. federal, State, and county agencies etc.) that the emergency has shif ted to a recovery phase. Recovery actions that plan for or may result in radioactive release will be evaluated by the Recovery Operations Manager and his staff as far in advance of the event as is possible. Such events and data pertaining to the release will be reported to the appropriate offsite emergency response organization and agencies. Am. 4 4-150 _ 1457 119

4.10 REFERENCES 4.10.1 Title 's, Code of Federal Regulations 4.10.1.1 Part 20, Standards for Protect ion Against Radiation 4.10.1.2 Part 50, Licensing of Production and Utilization Facilities 4.10.1.3 Part 50, Appendix E, Emergency Plans for Production and Utilization Facilities. 4.10.1.4 Part 100, Reactor Site Criteria 4.10.2 USNRC Order and Notice of Hearing, Docket No. 50-289, dated August 9, 1979 4.10.3 Section 13.3, Emergency Planning, of Regulatory Guide 1.70, Revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants 4.10.4 Regulatory Guide 1.101, Revision 1 dated March 1977, Emergency Plau.ing for Nuclear Power Plants 4.10.5 NUREG-75/087, Revision 1, USNRC Standard Review Plan 10.5.1 Sect ion 9.5.1, Fire Protection Program 10.5.2 Section 13.3, Emergency Planning 4.10.6 Regulatory Guide 1.97, Revision 1 dated August 1977, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident 4.10.7 US Environmental Protect ion Agency Manual EPA-520/1-75-001, September 1975, Manual of Protective Action Guic'es and Protective Actions for Nuclear Incidents 4-151 _ 1457 120

4.10.8 NUREG-0396, EPA 520/1-78-016, November 1978, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants 4.10.9 Federal Radiat ion Council Report No. 7, May 1965, Background Material for the Development of Radiation Protection Standards 4.10.10 National Council on Radiation Protection Report No. 39, January 15, 1971, Basic Radiation Protection Criteria 4.10.11 ANS-3.2/ ANSI N18.7-1976, Administrative Controls and Quality Assurance for the Operations Phase of Nuclear Power Plants 4.10.12 Regulatory Guide 1.120 (for comment), Revision 1 dated November 1977, Fire Protection Guidelines for Nuclear Power Plants 4.10.13 Three Mile Island Nuclear Station Operating License No. DPR 50 (Docket No. 50-289), including: 4.10.13.1 Appendix A, Technical Specifications (Safety) 4.10.13.2 Appendix B, Technical Specifications (Environmental) 4.10.14 Regulatory Guide 1.16, Revision 4 dated August 1975, Reporting of Operating Information - Appendix A, Technical Specifications 4.10.15 NUREG-0578 (extracts), July 1979, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations 4.10.16 NUREG-0600 (extract s), Investigation Into the March 28, 1979 Three Mile Island Accident by Office of Inspection and Enforcement, dated August 1979 4.10.17 Three Mile Island Nuclear Station Final Safety Analysis Report Am. 4 4-152

                                                                                   \457 \2\

4.10.18 USNRC Information Report SECY-79-450, dated July 23, 1979, Action Plan for Promptly Improving Emergency Reparedness 4.10.19 NRC Emergency Planning Review Guideline Number One - Revision One - Emergency Planning Acceptance Criteria for Licensed Nuclear Power Plants, dated September 7, 1979 4.10.20 NUREG-0610, USNRC Draft Emergency Action Level Guidelines for Nuclear Power Plants, dated September 1979 4.10.21 Proposed rule change to 10 CFR 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, published in the Federal Register on Augost 29, 1979 4.10.22 Commonwealth of Pennsylvania Disaster Operations Plan, Annex E, Emergency Nuclear Incidents (Fixed Nuclear Facility), dated September 1979 4.10.23 Department of Environmental Resources, Bureau of Radiation Protection Plan for Nuclear Power Generating Station Incidents, dated September 1979 4.10.24 Dauphin County Of fice of Emergency Preparedness, Act.an and Response Plan for Emergency Personnel and Citizens, dated April 1979 4.10.25 York County Evacuation Plan for the Three Mile Island Nuclear Power Plant, dated April 1979 4.10.26 Lancaster County Emergency Evacuation Plan, dated April 4, 1979 4.10.27 Cumberland County Of fice of Emergency Preparedness, Evacuation Plan 79-1 For Response to a Nuclear Accident at Three Mile Island Nuclear Station 4-153 Am. 4 1457 122 7

4.10.28 Lebanon County Emergency Managemen: Agency, Emergency Operations Plan, Radiation Incidents - Three Mile Island Nuclear Power Plant, dated April 11, 1979 4.10.29 Metropolitan Edison Company's Emergency Communications Plan for Three Mile Island Nuclear Generating Station 4.10.30 Three Mile Island Nuclear Station Procedures 4.10.30.1 Administrative Pro.cedures 4.10.30.2 Health Physics Procedures 4.10.30.3 Emergency Procedures

4. 10.30.4 Security Procedures

4. 10.30.5 Alarm Response Procedures 4-154 Am. 4 M,

i457 123

FIGURES AND TABLES 1457 124 4-155 Am. 4

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TABLE 1 MAJOR RESIDENTIAL AREAS POPULATION CENTERS GREATER THAN 500 WITHIN A 5-MILE RADIUS OF THE SITE AREAS POPULATION DISTANCE AND DIRECTION FROM SITE Goldsboro 576 1.5 miles W York Haven 671 2 miles S Royalton 1040 2.5 miles N POPULATION CENTERS GREATER THAN 2000 WITHIN A 10-MILE RADIUS OF THE SITE ARIAS POPULATION DISTANCE AND DIRECTION FROM SITE Middletown 9080 4 miles N Steelton 8555 6 miles hn Elizabethtown 8072 6 miles ESE Highspire 2946 7 miles NW POPULATION CENTERS GREATER THAN 10,000 WITHIN A 40-MILE RADIUS OF THE SITE ARIAS POPULATION DISTANCE AND DIRECTION FROM SITE Harrisburg 68,061 12 miles NW York 50,335 14 miles S Columbia 11,237 15 miles SE Lebanon 28,572 20 miles NE Lancaster 57,690 25 miles ESE Carlisle 18,079 27 miles W Hanover 15,623 29 miles SSW Note: Population approximate. 1457 \4e3

                                                                            ^ * '

4-176

TABI.E 2 Scil 001,S I,0CATED WITillN A 10 MILE RADillS OF THI SCIL 001, AREA EMPl.OYMENT ENROI.lllENT f.lizabethtown Area Public Schools Elizabethtown 272 3720 f.ower Daugdiin School District lia r r i sburg 410 4092 Milt on llershey School flershey 840 1448 Bishop McDevitt Iligh School llarrieburg 77 1205 liarrishurg School Dist riet lia r r i sbu rg i110 11848 flishlletowe Area School District Hiddletown 276 3140 St eelt on-Ilighspire School Dist rict St ee t t on-liigh spi re 150 2l00 West Shore School District New Cumberland 825 10417 Northeast York Count y School be st rict Mt .Wol f 253 3270 St . John's Elementary Enhant 4 109 St . Peter's Elementary Elizabethtown 4 106 1101 y Fami l y Element.:ry lia r r i shu rg 9 305 7 18o1y Name of Jesus llarrishurg II 393 tj Dur I ady of t he Blessed Sac rament lia r r i s bu rg 10 309 Sacreet lleart llarrisburg 7 250 St, Catherine Laboure llarrisburg Il 270 St . Francis of Assisi Ilarrisburg 8 288 The Cathederal School liar r i sburg 8 184 St. Margaret Mary lia r r i sburg 17 493 St, Joan of Arc Element ary lle r she y 11 328 Seven Sorrows B.V.H. Element ary  !!iddletown 10 246 St. Theresa Elementary New Cumberland 14 341 Assumpt ion of the B.V.H. Elementary St eet ton 8 158 Bishop Neuman StceILon 9 193 - St . Peter's St ecit on 4 81 A Penn St at e IJnivers k t y (Capit ol Campus) Mid.Iletown 248 2006 (.y1 Hilt on liershey Medl cal Center fle r shey 1123 300 N Elizabethtown College Elizabethtown 305 1732 IIarrishurg Arca Co$munit y Col lege lia r r i shurg 250 4400

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TABLE 1 Il0SPITALS LOCATED WITillN A 10 MILE RADillS OF TH1 Il0SPITAI. AREA EMPl.OYMENT

• PATIENTS Commun i t y General Ost eopat hic Ilos pi ta l lia r r i sbu rg 400 180 Danphin County llospita1 lia r r i s bu rg l

500 t 500 liar risburg Ilospital liarrisburg 1668 666 liarrisburg Polyclinic Ilospital liarrisburg 1300 700 liar r i sburg St at e Ilos pit al lia r r i s bu rg 1060 1281 Milton llershey Medical Cent er IIe r shey 1123 350 Elizabethtown State Ilospital for Crippled Children Elizabethtown 250 100 t = a 1 i i M M LD N M N Am. 4

TABLE 4 PERCE!C OF LAND USED BY COUNTY USE DAUPHIN YORK LANCASTER Forest 6 Woodland 45.2 22.6 13.7 Crops 31.9 49.8 62.5 Pasture 4.8 10.1 9.4 Urban 8.6 6.5 7.8

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TABLE 6

 .                           ACCIDENT CLASSIFICATIONS ACCIDENT                              EMERGENCY CLASS

1) Uncompensated Operating Reactivity Changes Unusual Event

2) Startup Accident Unusual Event

3) Rod Withdrawal Accident at Rated Power Operation Unusual Event

4) Moderator Dilution Accident Unusual Event

5) Loss of Coolant Flow Unusual Event

6) Stuck-Out, Stuck-In, or Dropped Control Rod Accident Unusual Event

7) Loss of Electric Power Unusual Event

8) Steam Lir e Failure Alert

9) Steam Generator Tube Failure Site Emergency

10) Fuel Handling Accident Site Emergency

11) Rod Ejection Accident Alert

12) Loss of Coolant Accident Site Emergency

13) Maximum Hypothetical Accident General Emergency

14) Waste Gas Tank Rupture General Emergency 4-181 Am. 4 1457 150

TABLE 7 INSTRUMENTS FOR ACCIDENT DETECTION ACCIDENT INSTRUMENTS FOR DETECTION RANCE

1. Uncompensated Operating 1. Reactor Average Temp. 520-620 F React ivity Changes 2. Hot Leg Temp. 520-620 F

3. Cold Leg Temp. 520-620 F

4. Power Range Monitor 0-125% psig

5. Reactor Coolant Pressure 1700-2500 psig

2. Startup Accident 1. Source Range Monitor .1-10 c/s

2. Reactor Average Temp. 520-620 F

3. Reactor Coolant Pressure 1700-2500 psig

4. Hot Leg Temp. 520-620 F

3. Rod Withdrawal Accident 1. Hot Leg Temp. 560-620 F at Rated Power Operation 2. Reactor Pressure 1700-2500 psig

3. Power Range Monitor 0-125%

4. Mcderator Dilut ion 1. Power Range Monitor 0-125%

Accident 2. Reactor Coolant Pressure 1700-2500 psig J. Equipment Status Lights (i.e. Valve position)

4. Pressurizer Level 0-400"

5. Loss of Coolant Flow 6

1. Total Reactor Flow 0-180x10 lbm/hr 6

2. Loop Flow . 0-90x10 lbm/hr

6. Stuck-out, Stuck-in or 1. Power Range Monitor 0-125%

Dropped Control Rod. 2. Reactor Average Temp. 520-620 F

3. Reactor Pressure 1700-2500 psig 4 Control Rod Posit ion Indication 0-100%

4-182 Am. 4 1457 151

TABLE 7 INSTRUMENTS FOR ACCIDENT DETECTION ACCIDENT INSTRUMENTS FOR DETECTION RANCE

7. Loss of Power 1. In plant bus voltmeters 0-9KV

2. Switchyard bus voltmeters 0-500KV

8. Steam Line Failure 1. Main Steam Pressure 0-1200 psig

2. Steam Generator Level 0-100%

3. Reactor Coolant Pressure 1700-2500 psig 4 Power Range Monitor 0-125%

9. Ste.sm Generator 1. Reactor Coolant Pressure 1700-2500 psig Tube Failure 2. Steam Generator Level 0-100%

3. Condenser Exhaust Atmospheric 10-10 cpm.

Monitor

10. Fuel Handling Accident 1. Fuel Handling Bridge Aux. .1-10 mR/hr Radiation Monitor

2. Fuel Handling Bridge Main .1-10 mR/hr Radiation Monitor 6

3. Fuel Handling Building Atmo- 10-10 cpm, spheric Monitor 0

4. Aux. & Fuel Handling Bldg. Stack 10-10 cpc.

11. Rod Ejection Accident 1. Power Range Monitor 0-125%

2. Reactor Coolant Pressure 1700-2500 psig

3. Pressurizer Level 0-400"

4. Reactor Building Pressure 0-100 psig 1457 152 4-183 Am. 4

TABLE 7 INSTRUMENTS FOR ACCIDENT DETECTION ACCIDENT INSTRUMENTS FOR DETECTION RANGE

12. Loss of Coolant 1. Reactor Building Pressure 0-100 psig

2. Reactor Coolant Pressure 1700-2500 psig

3. Reactor Building Hi Range .1-107 mR/hr X100 Radiation Mon 'or 6

4. Reactor Buildir.g Stack Atmo- 10-10 epm.

shperic Monitor

13. Maximum Hypothetical 1. Reactor Building Pressure 0-100 psig 6

2. Reactor Building Stack 10-10 cpm.

Atmospheric Monitor

3. Reactor Building Hi Range Radiation Monitor

14. Waste Gas Tarf. Rupture 1. Area Gamma Monitor - .1-10 mR/hr Auxiliary Building Entrance Elevation 305 ft.

2. Auxiliary Building Atmospheric 10-10 cpm.

Monitor 6

3. Waste Gas System Exhaust 10-10 cp,,

1457 l'3 4-184

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TABLE 9 RADIATION MONITORING SYSTEM -

SUMMARY

DESCRIPTION INSTRUMENT RAb1ATION MONITOR INSTRUMENT SETPOINT CIIANNEL _ _ _ _ _ _ _ _ DESCRIPTION _ _ RANGE Alert"-~~ lii gh ~ _I NTE R LOC _K_S Area Radialinn Monitors: C1 Control Room .1-10 mR/hr G2 Radiochemical I,ab .1-10 mR/hr G3 Sampling Roo.a .1-10 mR/hr G4 Ilo t Machine Shop .1-10 mR/hr G5 Reactor Building per ACC Door .1-10 mR/hr G6 Fuel llandling Bridge Aux. .1-10 mR/hr G7 Fue1 IlandIing Bridge Main .i-10 mR/hr G8 Reactor Building liigh Range .1-10 mR/hr x 100 v 8 y G9 Fuel llandling Bridge .1-10 7 mR/hr

• Pefueling Building GIO Aux. Bldg. Entrance, El. 305' .1-10 mR/hr Gil Waste Decay Tanks .I-10 mR/hr G12 Solid Waste Area .1-10 mR/hr Atmospheric Radiation Monitors:

Al Part iculate-Cont rol Tower 10-10 cpm 6 Al lodine-Control Tower 10-10 c p, 6 Al " Gaseous-Control Tower 10-16 cp, W *

                             -J
                           ~                                                                                                          Am. 4 W

W

TABI.E 9 RADIATION HONITORING SYSTEH -

SUMMARY

DESCRIPTION INSTRIIMENT RADIATION MONITOR INSTRUMENT SETrolNT CIlA_f!NEL_________ DESCRIPTION _. _ _ _ . _ _

                                                              .-RANGE..----      Alert  .-- -- . - -- .Il i gh ~

INTERI,0CKS Atmospheric Radiation ifonitors: A2 Part iculate-Reactor nuiIding 10-10 cpm A2 lodine-Reactor Building 10-10 cpm A2 Caseous-Reactor Building 10-10 cpm A4 Part iculate-Fuc t Iland ling Bldg. 10-10 cpm A4 lodine-Fuel lland ling Bldg. 10-10 cpm A4 Gaseous-Fuel lland ling Bliig. 10-10 cpm A5 Condenser Exhaust 10-10 cpm Y 6 g A6 Part iculate-Auxiliary Building 10-10 cpm 6 A6 lodine-Auxiliary Building 10-10 cp, A6 Gaseous-Auxiliary 11oilding 10-10 cpm A7 Waste Gas System Exhaust 10-10 ' c pm 4 g A8 Par t icu la t e-Aux . & Fuel llandling y Building Stack 10-10 cpm - A8 lodine-Aux. & Fuel llandiing 6 Ur, Bldg. Stack 10-10 cpm Ch A8 Gascous-Aux. & FucI Ilanilling 6 HIdg. Stack 10-10 cpm Am. 4

TAlu.E 9 RADIATION FlONITORING SYSTEM - SUPfMARY DESCRFPTION INSTRilHEffr RADIATION HONITOR INSTRilMENT SETPOINT CllANNEl. DESCRIPTION RANGE Alert lligli INTERI,0CKS At mo s pl e r_ic_ Rad _i n t i on Hon i t o r s ( Con t ' d_) : A-9 Part icula te-Reac t or 6 Iloilding Stack 10-10 cpm A-9 Iodine-Reactor linilding St ack 10-10 cpm A-9 Ga s eou s-Re ac t o r Building Stack 10-10,7 cpm Li_qu i d Rad i n t i_on_tfon i t or s : LI-III Primary Coolant I.etdown 10-10' cpm Ll-I,0 Primary Coolant Letdown 10-10' cpm s 1 5 cn L2 6 Decay llent Closed Cycle, Loop A 10-10 cpm I.3 6 Ikcay llent Closed Cycle, Loop B 10-10 c p, IA 6 4 Nuclear Services Closed Cycle 10-10 cpm LJ1 L5 Spent Fuel 10-10 cpm LJ7 N 6 L6 Radioact ive Waste Discharge 10-10 cpm I.7 Radioactive Waste Discharge 10-10' cpm Am. 4

TABLE _9, RADIATION HONITORING SYS_ TEM -

SUMMARY

DESCRIPTION INSTRUMEtri RADI ATit"; i;'NITOR INSTRUMENT SETPOINT_.____. CIIANNEL DESCRIPTION RANGE Alert Iligh INTEstLOCKS L_iquid Radiation Honitors (Cont 'd ): L9 Intermediate Cooling Water 10-10 cpm 2x10 cpm 4x10 cpm T m e A

    'J1 N

s Am. 4

      ,y CO

TAnl.E 10 REMP - SAffPl.E COLI.ECTION AND ANALYSIS METHODS Sample Sampiing Sampie Size '% cedure Analysis Medium Method Collected Abstract Cross alpha Ap quareterly composit e of 13 weeks of filters sample is leached with nit ric weekly, cont innous air filters per sampling "" "'" ' " " ' ' ""E"'" " sampling through filter site (approx. 3600 ft3) onto planchette low level gas paper. flow proport ional count ing. Cross beta AP continuous weekly I filter low level gas flow propor t ional air sampling through (approx. 280 N3) counting filter paper RW,SW according to sampling 4 liters sample is evaporated, residue site, various compos it ing t rans ferred to planchot te, and frequencies act ivity measured by low level counting n i Gamma Spect roscopy AP monthly and quarterly 4 weeks or 13 high resolut ion Cel.i gamma $ composites of weekly, weeks of filt ers isotopic analysis continuous air sampling (approx. 1100 or3600 H3) through filter paper AI continuous weekly air I cartridge same sampling through filter (approx. 280 N ) H grab sample according 8 liters same RW, SW to sampling site, various 4 liters same compo s i t i ng frequencies AQF,AQp grab sample 2 kg name AQS,FPL,FPF ( Tl N tm. 4 4

TABLE 10 REMP - SAMPLE COLLECTION AND ANALYSIS HETilODS Sample Sampling Sample size Procedure Analysis Medium Method Collected Abstract 11 - 3 H,SW according to sampling 4 liters Water is converted to site, various compositing hydrogen, Methane added frequencies and counted in I liter proport ional coimter. 1-131 H,SW grab sample for H and 8 liters anion-exchange, s o l ve n t. according to sampling extraction, palladium iodide site, various compositing precipitate, low level gas frequencles flow count ing Sr-89,90 Ap quarterly composite of 13 weeks of filters Strontium in sample (with weekly, cont inuous air carrier) is precipit at ed as sampling through filter per sampling (approx. 3600 sity) M SrNO mount, Sr-70 inferred 3 paper from Y-90 on yttrium 4 oxalate mount, low level y gas flow counting AQF grah sample 2kg similar AQS grab sample 2kg similar according to sampling

                                                                   ~

RW,SW 4 liters similar site, various compo-siting frequencies M grab nample 8 liters oxalate precipitatic of TCA filtrate, barium and iron scavenge, 7 day yttrium ingrowth, Sr-m on yttrium oxalate mount, w low level gas flow count ing.

         ' 51 TLD     N      ID    quarterly exposure                TLD                 Thermoluminscent dosimetry h

CD Am. 4

TABLE 11 CONTkOL ROOM INSTRUMENTATION INSTRUMENTATION / INDICATION INSTRUMENT eiSTEM DESCRIPTION RANCE NUCLEAR 1. Power Range Monitor 1. 0-125%

                                                                  ~
                                                                            ~3 IISTRUMENTATION      2. Intermediate Range Monitor           2. 10     to 10     amps 6

3. Source Range Monitor 3. 0.1 to 10 ep, REACTOR 1. Hot Leg Temp (Th) 1. 520-620 F COOLANT 2. Pressurizer Temperature 2. 0-700 F

3. Cold Leg Temp (Tc) Wide 3. 50-650 F

4. Cold Leg Temp (Tc) Narrow 4. 520-620 F

5. Reactor Coolant Average Temp (Tave) 5. 520-620 F

6. Reactor Coolant Unit Ave 6. 0-70 F Differential Temp ( T)

7. Pressurizer Liquid Level 7. 0-400"

8. Reactor Coolant Pressure (Wide) 8. 0-2500 psig.

9. Reactor Coolant Pressure (Narrow) 9. 1700-2500 psig.

10. Reactor Coolant Total Flow 10. 0-180 x 100 lbm/hr

11. Reactor Coolant Loop Flow 11. 0-90 x 106 lbm/hr

12. Equipment Status Lights MAIN STEAM 1. Steam Generator Level 1. 0-100%

2. Main Steam Pressure 2. 0-1200 psig

3. Valve Position Indication 3. Not Applicable FEEDWATER 1. Feedwater Flow 1. 0-6x106 lbm/hr

2. Equipnent Status Lights 2. Not Applicable EMERGENCY FEEDWATER 1. Steam Pressure to Emergency 1. 0-500 psig Feed Pump Turbinc

2. Emergency Feed Pump Turbine Speed 2. 0-5000 rpm

3. Emergency Feed Pump (Steam-driven) 3. 0-1500 psig Discharge Pressure

4. Emergency Feed Pump (Motor-driven) 4 0-1500 psig Discharge Pressure

5. Equipeent Status Lights 5. Not Applicable Page 1 of 2 Am. 4 4-192 1457 161

TABLE 11 CONTROL ROOM INSTRUMENTATION INSTRUMENTATION / INDICATION INSTRUMENT SYSTEM DESCRIPTION RANGE MAKEUP & PURIFICATION 1. Makeup Tank Level 1. 0-100"

2. High Pressure Injection Flow 2. 0-500 gpm

3. Equipment Status Lights 2. Not Applicable CORE FLOODING 1. Core Flood Tank Level 1. 0-15 ft.

2. Core Flood Tank Pressure 1. 0-800 psig REACTOR BUILDING SPRAY 1. Reactor Bldg. Spray Flow 1. 0-1800 pgm

2. Sodium Thiosulfate Tank Level 2. 0-50 ft.

3. Sodium Thiosulfate Tank Temperature 3. 0-200 F

4. Sodium Hydroxide Tank Level 4. 0-50 ft.

5. Soduim Hydroxide Tank Temperature 5. 0-200 F

6. Reactor Bldg. Pressure (Wide) 6. 0-100 psig

7. Reactor Bldg. Pressure (Narrow) 7. 0-10 psig

8. Equipment Status Lights 8. Not Applicable DECAY HEAT REMOVAL 1. Borated Water Storage Tank Level 1. 0-60 ft.

2. Decay Heat Removal Pump Flow 2. 0-5000 gpm

3. Low Pressure Injection Flow 3. 0-5000 gpm

4. Equipment Status Lights 4. Not Applicable REACTOR BUILDING 1. Reactor Bldg. Emergency 1. 0-100 psig EMERGENCY COOLING- Cooler A, B, & C RIVER WATER Cooling Water Pressure

2. Equipment Status Lights 2. Not Applicable 4-193 Am. 4 1457 162

TABLE 12 _P_E_R_ IODIC TRAINING OF EMERGE 5CY RESPONSE PERSONNE_L_ PcraonneMa[cjo'r[~~~ ' []- ~~_^ _~~[ ~~ "'~'[7r~di Ii tigd:Id 'l r'eqtIc~nIy' lnvoIved[P[r' sonne 1 _ _ ' ~ ~ ] } _ _~ [ ~ _ _ _)) Licensed Operators All Senior Reactor Reactor Operators and Senior Reactor 6:perators receive extensive periodic Operators and on-the-job and continuous formal trais.ing as scheduled and conduct ed by the henctor Operators operator requalification training program. This program shall include a comprehensive review of the THI Emergency Plan and Implement ing Document . Per :onnel respon- Duty Sect ion Super- 1he listed individuals are responsible for kne.eledge of the current THI sible for accident intendents, Eme rge nc y Emergency Plan, implement ing Document , Technical Specifications, and assessment and/or Director, Technical other related <;tation programs, plans, and procedures. The l i s t e.1 accident management. Su ppo rt and Ope ra- indviduals attend at least one meet ing per year to receive t raining on s t ions Support Cent e r the THI Emergency Plan and Implement ing Document . Detailed inst ruct ions 3 Coordinators, Radio- with special at tent ion given to the use of the done project ion techniques 4 I logical Assessment will also be provided, y Coordinator, Radio-logical Analysis Support Engineers, Operations Coordinator, Emergency and Assis-tant Emergency Sup-port Directors, Group Leader llealth Physics / Chemi st ry Support , Environment al Assens-ment Coordinator, Group Leader Technical __, Support b LD N ___ _ _________ _ _ _ _ _ ___.__ ___ ________ -____________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ U Am. 4

e _ t h d _ l e c _ 4

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TABLE 12 PERIODIC TRAINING OF EMERGENCY RESPONSE PERSONNEL

       ~

Personn W k y g[o}r [ _ ____I n voTv'ed Ee'r' son ~n e l

                                                                                       ]'[_ ] ] '_ [ _ 7 r'aini[g] S y Yr'e'qSe~n~c'y~ ~ ] ] '_ ^][ _ _
                                                                                                                                                                                    ~

Emergency Hepair Designated Maint en- The training program for Emergency Repair Team personnel will i nc l eule , Imt Team (s) ance Depart ment sh i f t not be limited to: responsibilities during an emergency, radiological workers considerations, ava!lahilit y of damage control equipment, use of commun-ication systems, and the interfaces with other emergency t eams. This training will be conducted at least annually. Site Security and Security Supervisor, The personnel listed to the lef t will receive training on at least Su ppo rt Secur i t y Security Coordinator, an annual basis. They *n turn will be responsible for training their Forces Security Sergeants, res pec t ive security of ficers and other support personnel. 'I b e training Group Leader-Securi ty program will include at least the following subject s: a review of the Support applicable part s of the THI Emergency plan and Implement ing Document

                                           ;                                           with emphasis on the classification of emergencies, commun ient ions ,

a l and spec i fic areas of responsiblity; personnel account abili t y; personnel 5

• i and vehicle access control during emergencies; evacuation control; and

                                           }                                            int erfaces with of f aite support organizations.

Fire Brigade Operations Conr- This training, which is provided at least on a semi-annual basis to d(nator, Sh i f t each person involved, i s g i ven by Met -Ed instructors trained in fire Supervisors and fighting. The program will include, but not be limited to t he t ypes Fo reme n , and assigned of fires and their part icular hazards, equipment to be used on each team members on each type of fire, the installed fire detect ion and prot ect ion systens, sh i f t . portable firefight ing equipment , res pi rat ory prot ect ion ilev ices , anil radiological hazards existing during fire emergencies. In aiblit ion, a review of firefighting procedures and techniques will be i nc linled in the training program. Furthermore, pract ical demonst rat ions of firefighting will be given which will include t he part ic ipal ion of Fire Brigade personnel, b t J1 N ON l Am. 4 LJ1 ' p

TABLE 13 EXAMDLES OF ONSITE LABORATORY EOL'IPMENT

         ' Item No.            Eouipment Manufacturer / Description

1. Perkins - Elmer Model 403 Atomic Absorption Spectrophotometer

2. Beckman Model DU-2 Spectrophotometer

3. Fisher Model 25 Gas Partitioner

4. Hewlet - Packard 4096 Channel Analyzer

5. Hewlet - Packard 9830 Mini Computer

6. Packard Tri-Carb Tritium Analyzer

7. Beckman Wide Beta Counter Am. 4 4-197 1457 1A6

TABLE 14 TYPICAL CONTENTS OF 100 EMPLOYEE SIZE FIRST AID KIT Item No. Description

1. Adhesive Bandages

2. Adhesive Tape - various sizes

3. Plastic First Aid Tape

4. Towelettes

5. Gauze Bandages - various sizes

6. Gauze pads - various sizes

7. A=m",ia inhalents

8. Cotton

9. Eye pads

10. Cotton swabs

11. Tongue depressors

12. Pencil and paper

13. Scissors 14 Tweezers

15. First Aid Guide Booklet 1457 167 Am. 4 4-198

TABLE 15 TYPICAL FIRST AID ROOM IN'."dNTORY Item No. Description

1. Ambulance - type litter

2. Cervical collars

3. Inflatable splints

4. Triangular bandages

5. Ammonia inhalants

6. Swabs

7. Adhesive bandages - various sizes

8. Bandage compresses

9. Cold pacs

10. Burn spray

11. Eye pads

12. Poison antidote kit

13. Gauze - various sizes 1457 168 4-199 Am. 4

TABLE 16 MISCELLANEOUS MEDICAL EOUIPMENT TYPICALLY ONSITE Item No. Description

1. Trama dressings

2. Oxygen

3. Saline solut ion 4 Hare Traction Splint

5. Laerdal Suct ion Unit

6. Examination table

7. Blankets and sheets

8. Ambulance

9. Eye wash

10. Sterile water

11. Blood pressure cuff and stethoscope

12. Splints 1457 169 4-200 Am. 4

TABLE 18 TYPICAL ONSITE DAMAGE C0hTROL EOUIPMEh7 Item No. Description

1. Small hand tool kit

2. Mauls and sledges

3. Hand saws 4 Welding machines

5. Cutting torches

6. Axes

7. Hatchets

8. Chisels

9. Wooden plugs

10. Pipe patching kits

11. Assorted clamps

12. Assorted turn buckles

13. Chain-falls 14 Hydraulic jacks

15. Ventilation blowers

16. Induct ive type ammeter

17. Multimeters

18. Rubber gloves

19. Pipe fittings

20. Submersible pumps

21. Carts

                                                   ]f}[ ][]

22. Sand bags 4-201 Am. 4

APPENDICES A THROUGH I These appendices are available on site for NRC review. 1457 171 4-202 Am. 4

  . p     b w a

0 9 9 9 9 i 9 4 N t

                              ,        e-'

r. 6, 9 ' N

5.0 THREE MILE ISLAND NUCLEAR STATION ORGANIZATION 5.1 GENERAL The direction of the Three Mile Island Nuclear Generating Station (TMINGS) is now under the control of the TMI Generation Group. This group was formed to take advantage of the wealth of nuclear experience represented by management and technical staff from within the GPU Service Corporation and Metropolitan Edison Com-pany. This reallignment more than tripled the number of profes-sionals that have TMINGS as their primary responsibility. The head of the TMI Generation Group is an of ficer of both the GPU Service Corporation (Vice President - Generation) and Metro-politan Edison Co. (Sr. Vice President). Therefore, he has the necessary control of, and access to, the resources of both com panie s . There are senior management personnel with average technical experience well over 20 years, reporting to the head of the TMI Generation Group in the areas of:

         . Operations
         . Technical Functions
         . Environment , Health and Safety
         . Reliability Engineering
         . Maintenance
         . Procurement, Contract Administration and Materials Receiving, Inspection Warehousing, Issuance and Account-ability (a site located group administratively responsible to the VP-Mtis. Management has been assigned by him to be functionally accountable and be dedicated solely to the Site's requirement and is to be measured for its respon-siveness to site requirement.)

Various steps have been taken to strengthen key function in the organization. Examples of this are:

1. The GPU Service Company and Metropolitan Edison Company, Quality Assurance and Control Organi-zations were merged and operating QA for TMI is their ma a or function.

2. The radiological control f unction for Unit-1 has been elevated so that it reports directly to the the Vice President - Nuclear Operations.

The following sections describe the pertinent details of the TMI Generation Group. 1457 173 5-1 Am. 4

5.2 STATION ORGANIZATION The Vice President, Nuclear Operations utilizes the following management staff in carrying out his responsibilities:

• Director - Technical Support Superintendent - Unit 1 Superintendent - Unit 2
• Manager - Training
• Manager - Support Services and Logistics The Three Mile Island Nuclear Station organization as shown in Figure 5.2.1 will function in five main areas: Unit 1 operations and preventative maintenance, Unit 2 oeprations and preventative maintenance, technical support, training, and support services including Health Physics, Chemistry, and Security.

The Operations Group under the Unit 1 Superintendent will be responsible for the day-to-day operation of the unit. Unit I will have a Shift Foreman directing the operations of each shif t through the Control Room operators and Auxiliary Operators. A maintenance force supporting TMI-l in the areas of electrical, mechanical and instrument control preventative maintenance and surveillance will also report to the Unit 1 Superintendent. This maintenance force will be suppliemented by additional forces under the Director of Maintenance - GPU for corrective mainte-nance assignments. The Technical Support Group under the Director - Technical Support will consist of lead engineers in such disciplines as nuclear, mechanical, electrical, and instrument and control engineering to whom other engineers and analysts assigned to TMI-I will report. In addition, Technical Engineers will be assigned on each operating shift to maintain technical liaison and coordination between operating shift personnel and the technical support engineering staff. The Training Department will function primarily in the three main areas of operator training, technician training and accelerated retraining of operators. The operator training section is organized to support both licensed operator and non-licensed operator training. The technician training section will support training of technicians in both the maintenance and health physics areas. The accelerated re-training program section is designed to present an augmented training program as a result of the TMI-2 accident. The Operations and Maintenance group for TMI-2 will be respon-sible for the day-to-day operations and preventive maintenance and surveillance of Unit 2. The Support Services and Logistics group will function in the areas of facilities, office management, personnel, station security and health physics and chemistry.

                                                         \451 \14 5-2                       Am.4

The following subsections detail the functions and responsibili-ties of various station supervisory personnel. 5.2.1 Vice President - Nuclear Operations

a. Function The Vice President - Nuclear Operations (Plant Manager) in carrying out his management responsibility for day-to-day plant operations directs the management staff, identified in paragraph 5.2 Statien Organization, in executing the following TMI-l operational functions.

Shif t daily operations and surveillance in accordance with Technical Specification requirements. Preventative and minor corrective maintenance on safety related systems and components. Administrative controls related to Technical Specification compliance not specifically assigned elsewhere. Radiological controls including compliance with Radiological Effluent Technical Specifications. Primary end Secondary plant chemistry. 4 Radioactive waste processing and treatment including shipment. Refueling operations and operational related shutdown sur-veillance requirements. Training of licensed and non licensed operations personnel, Radiation / Chemistry technicians and maintenance personnel. Planning of day-to-day maintenance, operations surveillance and refueling activities. Plant engineering support of maintenance requirements and shif t operations, including liaison with the of fsite Tech-nical Functions Group.

b. Responsibility The Vice President - Nuclear Operations in carrying out his responsibility for overall direction of day-to-day TMI-l operations is responsible for:

TMI-l Technical Specification and Regulatory Requirements compliance , (unless sspecifically assigned elsewhere , e.g. , the Radiological Environmental Monitoring Program and non-Radiological Environmental Monitoring Program) through the direction of the manages identified in paragraph 5.2.

                                                                  \457 175 5-3                    Am. 4

Direction of the TMI Unit 1 Superintendent in the execution of his responsibilities which are set forth in paragraph 6.1 of the TMI-l Technical Specifications. TMI-l compliance w;th all commitments made in the TMI-l restart report submitted in response to the TMI-1 shutdown order dated 9 August, 1979 unless specifically assigned elsewhere. Developing and implementing the Security Plan, Fire Protection Plan and Radiation Emergency Plan. Compliance with the company approved Quality Assurance program, the Security Plan,. the Radiation Emergency Plan, the Fire Protection Plan and the Radiation Protection Plan. Insuring that training programs for licensed operators, non-licensed operators, maintenance personnel, Shift Technical Advisors, Radiation Chemistry Technicians and the Security Force are implemented and maintained in accordarce wita regulatory and management requirements. Direct Interface with the Directors identified in paragraph 5.3 to insure the necessary TMI-l support is provided in the areas of Engineering Change Modification coordinated review and approval, major corrective maintenance and esnstruction, Nuclear Safety Analysis, quality assurance, licensing, environmental programe, Licensee Event Report Review, labora-tory analysis and Generation Review Committee support in accordance with the TMI-1 Technical Specifications. Insuring that appropriate management and administrative control systems and procedures are developed, implemented, and compiled with as necessary to fulfill the other listed responsibilities. Evaluation of the adequacy of the Station Staf f in terms of number assigned and qualifications.

c. Authority The Vice President - Nuclear Operations has the authority to:

Implement the Radiation Emergency Plan. Order the shutdown and cooldown of TMI-1 whenever the health and safety of the public is endangered. Initiate emergency procurement. 1457 176 5-4 Am. 4

d. Minimum Qualifications The Vice President - Nuclear Operations shall possess as a minimum the qualifications outlined in ANSI 18.1, Section 4.2.1, Plant Manager. The Vice President (Plant Manager) shall have ten years of responsible power plant experience, of which a minimum of three years shall be nuclear power plant experience. A maximum of four years of the remaining seven years of experience may be fulfilled by academic training on a one-for-one time basis. To be acceptable, this academic training shall be in an engineering or scientific field generally associated with power production. The Vice President (Plant Manager) shall have acquired the experience and training normally required for examination by the NRC for a Senior Reactor Operator's License whether or not the examination is taken.

e. Incumbent Qualifications The incumbent received a Bachelor of Science Degree in Marine Engineering in 1960 from the U.S. Naval Academy. From 1960 to 1967, he spent six years on conventional destroyer ships in various capacities including Weapons Officer and Chief Engineer. In addition to the shipboard assignment, he epent one year at the U.S. Naval Nucleer Power School and qualified as Engineering Of ficer of the Watch at the DIG Prototype at We st Milton, New York. From May 1967 to September 1967 the incumbent was the Assistant to the Operations Supervisor at Yankee Atomic. In September 1967, he became Staff Engineer at the Saxton Nuclear Experimental Station where he remained until August 1968 when he assumed the position of Supervisor of Operations and test for Saxton. At that time he acquired a Senior Reactor Operator's License. In May 1970, he assumed the position of Supervisor of Reactor Plant Services at Saxton. In August 1970, the incumbent was transferred and assigned as Station Engineer at TMI. In that capacity, he was responsible for instrumentation, eler.trical, mechanical, nuclear, health physics and chemistry, site engineering and technical supervision. He remained in this position until January 1973, when he assumed the position of Assistant Superintendent of TMI. He acquired a Senior Reactor Operator's License in Feburary,1974. From January 1974, to June 1975, he held the position of Superintendent Nuclear Generating Station at TMI. In June 1975, he assumed the position of Manager - Generation Operations Nuclear in which he was responsible for day-to-day direction and Supervision of TMI.

In September 1976. he assumed the position of Manager - Generation Operations in which he was responsible for nuclear, fossil and hydroelectric generation. In May 1977, he assumed the position of Vice President - Generation and in August 1979, he was assigned to TMI as Vice President - Nuclear Operations and is presently serving in that capacity.

                                                        \451      \ll 5-5                     Am. 4

f. Interfaces The Vice President - Nuclear Operations, (Plant Manager) reports to and is held accountable for TMI-l operations by the Met-Ed Senior Vice-President.

In carryia; out his management responsibility for day-to-day IMI-l operations the Vice President of Nuclear Operations interfaces and communicates with the Directors shown on Figure 5.2.1 who also report to the Met-Ed Senior Vice President. The offsite Directors who interface and communicate with the Vice President of Nuclear Operations and their corresponding direct support responsibilities relative to TMI-l are identi-fied in paragraph 5.3. The Vice President of Nuclear Operations interfaces and communicates directly with the Director TMI-2 Reccvery and the TMI-2 Site Operations / Waste Management Manager to insure the separation, and independence of TMI-2 decontamination and restoration activities from TMI-1. Auditionally, this interface and related communications insures TMI-l installed waste handling equipment which is required for operation as described in the TMI-l FSAR is not relied on by operations at . TMI-2. The Vice President of Nuclear Operations has the ability to call upon various components of the Technical Functions and Quality Assurance Groups for assistance in the areas of process computers , Safety Analysis , f uel performance and f uel management. 5.2.2 Unit Superintendent

a. Funciton The Unit Superintendent reports directly to cha Vice President-Nuclear Operations and assists him in the overall operation and maintenance of the unit,

b. Responsibility This position has direct responsibility for operating the unit in a safe, reliable and efficient manner; is responsible for off-site radioactive discharges and the protection of personnel from radiation exposures; bears the responsibility for compliance with the oeprating licenses and the rules and regulations of the Commonwealth of Pennsylvania; supervises the operations group and Preventative maintenance group.

1457 178 5-6 Am. 4

c. Authority The authority of the Unit Superintendent, delegated by the Vice President - Nuclear Operations, is inherent in the position and commensurate with the assigned responsibilities.

It includes the authority to issue procedures, orders, and other directives required in the execution of the assigned responsibilities. Necessarily included is the responsibility for plant operation, compliance with Technical Specifications, radioactive discharges and personnel protection from radia-tion exposure is the authority to assign and prioritize requirements to the Technical Support, Training and Support Services and Logistics Groups. Similarly, the authority of the Unit Superintendent includes the initiation, reprioriti-zation or cancellation of corrective maintenance, preventa-tive maintenance or construction in the execution of his responsibilities. All personnel within the confines of Unit I protected area are subject to the authority and direction of the Unit Superintendent. The Unit Superintendent may delegate his authority and share his responsibilities with the Supervisor of Operations or Shif t Supervisor during absences. This delegation of authority extends to the issuance of standing orders and directives in support of the responsibilities assigned. In the absence or incapacitation of the Vice President, Nuclear Operations, the Senior Unit Superintendent is delegated the authority of that office for the centralized control supervision, coordination and planning of all aspects of IMI Operations.

d. Minimum Qualifications The Unit Superintendent shall have a minimum of eight years of responsible power plant experience of which at least three years will be in nuclear power plant design, construction, startup, operation, maintenance, or technical services. A maximum of two years of remaining five may be fulfilled by academic training. The Unit Superintendent must hold a Senior Reactor Operat or license.

e. Incucbent Qualifications Unit 1 High School graduate. U.S. Merchant Marine Acadamy, Kings Point, NY (1963)

(Marine Engrg) 1965-70 Newport New Shipbuilding (assigned to various new const. nuclear submarines). 1970-71 Ne wport News Shipbuilding (Asst. Project Mgr.) for completion of reactor plant). 1971-73 Ne wport News Shipbuilding & Dry Dock Co. (Senior Test Supervisor) 2/73 to 8/74 GPU Corp. Supt.-Test Design. 5 years experience Nuclear Power Plant Operation ) h 5-7 Am. 4

f. Interfaces

1. Offsite The Unit Superintendent interfaces with company, corporate.

local, commonewalth, and federal government organizations in fulfillment of responsibilities assigned, state and federal regulations and directives received.

2. Other Unit In execution of his responsibilities, the Unit Superin-tendent maintains close liaison with the Unit 2 Superin-tandent to ensure compliance with Technical Specifications and other directives issued and to enhance safety and ef ficiency in operation of both units. In compliance with this directive, personnel subordinate to the Unit Superintendent maintain a similar liaison with counter-parts in Unit 2. This is emphasized in the operations area through the Supervisors of Operations and Shif t Supe rvisors.

5.2.3 Supervisor of Operations

a. Function .

The Supervisor of Operations has the respcnsibility for directing the actual day-to-day operation of the unit. He resports directly to the Unit Superintendent. The Supervisor of Operations coordinates operations, related maintenae.ce activities with the Supervisor of Maintenance, and Supervisor of Preventative Maintenance.

b. Responsibility This position is responsible for the day-to-day administration and direction of the Operations personnel, ensuring that the conditions of the plant oprating license are met, ensuring compliance with the technical specifications, administer the provisions of the management-union contract and execution of Company policy.

c. Authority The Supervisor of Operations Unit I has the authority and responsibility to order the plant shutdoan when in his judgment the safety of the plant or public is being compro-mised.

1457 180 5-8 Am. 4

d. Minimum Qualifications The Supervisor of Operations will have a minimum of four years of responsible power plant experience of which at least one year will be in nuclear power plant design, construciton,-

startup , operations, maintenance , or technical services. A maximum of two years academic or related training may be included as part of the remaining five years of power plant experi*nce. The Supervisor of Operations shall hold a Senior Reactor Operators License,

e. Incumbent Qualifications High School Graduate. Snowden Twp. , Lebrary, PA (College Prep.i U.S. Navy 12/60 to 11/68. Rank: ETI E-6, Electronic and Nuclear Jobs. 11 years experience Nuclear Power Plant Operation TMI and Saxton Experimental.

f. Interfaces

1. Offsite The Supervisor of Operations has no of fsite interfaces or responsibilities other than routine day-to-day plant operations related inquiries.

2. The Supervisor of Operations has n- technical responsibi-lities in Unit 2.

5.2.4 Shift Supervisor

a. Function This position directs the activities of the Shif t Foreman on his shif t and is cognizant of maintenance activities being performed while he is on duty.

b. Re sponsibility The Shift Supervisor is responsible for the broad perspective of station operations during his assigned shift and he reports directly to the Supervisor of Operations (NUREG 0578-Section 2.2.1.a).

c. Authority The Shift Supervisor has the authority and obligation to shut down the unit if , in his own judgment , conditions warrant this action and is responsible for the station during emer-gency situations from the Control Rom until relieved.

1457 181 5-9 Am. 4

d. Minimum Qualifications Each Shif t Supervisor shall have a high school diploma or an equivalent education. He shall have a minimum of 4 years power plant experience of which a minimum of one year will be nuclear power station operations or maintenance. A maximum of two years of academic or related education may be included as part of the remaining three years of required plant expe rience . The Supervisors in this category should hold a Senior Reactor Operator's License.

e. Incumbent Qualifications Incumbent A Righ School Graduate. U.S. Army 9/58 to 9/60. 10 years experience Nuclear Power Plant Operation - TMI.

Incumbent B digh School Graduate. 10 years experience in Nuclear Power Plant Operation - TMI. Incumbent C High School Graduate. U.S. Air Force. 9/59 to 9/63, Rank A/lc 10 years experience Nuclear Power Plant Operation-TM I . Incumbent D Righ School Graduate. U.S. Navy ET1 3/66 to 12/71. U.S. Navy Elec. (Electro) & Nuclear Power School (1964), Steel Valley Tech. Nuclear Power School (1965) (Electronics). (7 years experience Nuclear Power Plant Operation TMI) . Incumbent E High School Graduate. U.S. Air Force, 10/65 to 2/69. Rank: SGT. USAF Maintenance School Aircraft Mechanic. 10 years experience Nuclear Power Plant Operation - TMI. Incumbent F Righ School Graduate. Utah State Univ. (Forest Management). U.S. Navy 3/66 to 2/73, Rank: E5. 6 years expreience in Nuclear Power Plant Operation - TMI. 1457 i82 5-10 Am. 4

f. Interfaces

1. Offsite The Shift Supervisor has na of fsite interfaces or respon-sibilities during normal operation.

2. TMI-2 The Shift Supervisor has no technical responsibilities in Unit 2.

5.2.5 Shift Foreman

a. Function He reports directly to the Shif t Supe rvisor. he directs the activities of the unit operators on his shif ? and is cognizant of all maintenance activities being performed while he is on duty,

b. Responsibility The Shift Foreman is responsible for the actual operation of the unit during his assigned shif t.

c. Authority The Shif t Foreman on duty has both authority and the obliga-tion to shut down the unit _if, in his judgment, conditions warrant this action.

d. Minimum Qualificat' e ns Each Shif t Foreman shall have a high school diploma or an equivalent education. He shall have a minimum of 4 years power plant experience of which a minimum of one year will be nuclear power station operations or maintenance. A maximum of two years of academic or related education may be inc19ded as part of the remaining three years of required plant expe rience. The Foreman in this category should hold a Senior Reactor Operator's License. l

e. Incumbent Qualifications Incumbent A High School Graduate. Special training in Security Police and Conputer Operator. Bolling Air Force Base , Washington, D.C. (11/63 to 1/68), Presidential Honor Guard. 10 years experience in Nuclear Power Plant Operation - TMI.

1457 i83 5-11 Am. 5

Incumbent B High School Graduate. U.S. Navy Nuclear Power School,1965 - 6 month course Nuclear Power Prototype Training,1965. Machinist Mate Class "A" School,1964, five month course, Navy Enlisted Submarine School,1964, ::4intenance Data Collection for Supervisors, 1969 --- Naval Experience 8/65 to 2/69 Machinist Mate 1st Class POLARIS submiarine USS Theodore Roo sev elt (SSBN600) 2/6s to 2/71. 8 years experience in Nuclear Power Plant Operation - TMI. Incumbent C High School Graduate. Prince Georges Community College, Suitland , MD. (Engrt ). Elec. Mate , Nuclear Power School and Nuclear Power Trainicg Unit. U.S. Navy from 2/68 to 2/71. 8 years experience in N.' clear Power Plant Operation - TMI. Incumbent D High School Graduate. Cleveland Institute of Electronics. 4 years experience in Nuclear Power Plant Operation - TMI. Incumbent E High School Graduate. Correspondence Course in Mechanical Drafting. U.S. Navy (Submarine service) 9/57 to 9/60. 3rd Class. 10 years experience in Nuclear Power Plant Operation-TMI. Incumbent F High School Graduate. U.S. Navy, Electronic Technician 2nd Class , 7/63 to 8/70. 81/2 years experience in Nuclear Power Plant Operation - TMI.

f. Interfaces

1. Offsite The Shif t Foreman has not of fsite interfaces or responsi-bilities during normal operation.

2. TMI-2 The Shift Foreman has no technical responsibilities in Unit 2.

1457 184 5-12 Am. 4

                                                                        /

5.2.6 Supervisor Preventative Maintenance

a. Function This position directs maintenance personnel who execute the Met-Ed Preventative Maintenance Program. This position develops, monitors and modifies the PM Program.

b. Responsibility The Supervisor of Preventative Maintenance reports to the Unit Superintendent and is responsible for organizing, modifying , and conducting preventative maintenance for the Unit. The Supervisor - PM has the responsibility to identify /

justify resources requisite to accomplish the PM Program.

c. Authority Commit manpower and materials to execute the Preventative Maintenance Program.

Commit contractor resources to the PM Program. Establish and/or modify the PM Program.

d. Minimum Qualifications The Supervisor of Preventative Maintenance shall have seven (7) years of responsible power plant experience or applicable industrial experience, a minimum of 1 year which shall be nuclear power plant experience. An Associate Degree in an Engineering or Scientific field is preferred and may be credited to the remaining 6 years of experience. The indivi-dual should have non-destructive testing familiarity, craf t knowledge , and an understanding of electrical, pressure vessel, and piping codes.

e. Incumbent Qualifications High School Graduate. Naval Courses (Various) U.S. Navy 3/58 to 6/62. Rank E5.

Westinghouse Electric Corp. , Saxton, PA (Elec. Tech.) 7/66 to 11/68. 9 years experience in Nuclear Power Plant Operation and Maintenance - Saxton Nuclear Experi. Station and TMI.

f. Interfaces

1. Offsite None.

1457 i85 5-13 Am. 4

2. Other IMI Staff Supervisor - Unit Maintenance - coordination of work

                                              - allocation of resources
                                              - feedback on program ef fectiv ene ss Supervisor - Operations         - scheduling of work Director - Engineering             program technical support /

evaluations

3. TMI-2 Unit 2 Preventative Maintenance Supervisor - coordination of PM Programs.

5.2.7 Director - Technical Support

a. Function The Director of Technical Support will report to the Vice President - Nuclear Operations and is responsible for the coordination of the technical engineering staff including the Nuclear Engineering , Mechanical Engineering, Electrical Engineering , Instrument and Control Engineering and Shift Technical Engineer.

b. Responsibility The position is responsible for assisting the Unit Superin-tendent in the technical engineering for TMI Unit 1 or 2 in order to ensure safe, efficient and continuous generation.

The incumbent ensures overall safety of Unit Operations through the review and evaluation of changes to procedures , systems and equipment in the light of their effect upon the FSAR, the CFR, etc. This position is responsible for the ef fective direction of lead engineers and their functional areas, ensuring that technical support is provided in deci-sions involving all aspects of the Unit's operation, schedul-ing and coordination of all aspects involved with and during plant ref ueling , supe rvising budget preparation and controll-ing expenditures to conform to the unit budget, recommending various personnel actions, and ef fectively assists in the coordination of communications between TMI and corporate and System Engineering groups.

c. Authority The Director of Technical Support has the authority to approve minor plant design and modification work. He has the authority to approve Purchase Requisitions for material, equipment, supplies and services in the Engineering Area at IM I . He has the authority to effect procedure changes 1457 186 5-14 Am. 4

through the Plant Operating Review Committee. Additionally , he shall provide engineering support for the maintenance group.

d. Minimum Qualifications The Director - Technical Support shall have 8 years in responsible positions related to power generation, of which one year shall be nuclear power plant experience. A Bachelor of Science Degree in an Engineering or Scientific field is preferred and may e credited to the remaining 7 years of experience. The individual should have non-destructive testing familiarity, craft knowledge, and an understanding of electrical, pressure vessel and piping codes.

e. Incumbent Qualifications B.S. Mechanical Engineering - Villanova University,1963.

1963 - Cadet Eng. - Reading. 1965 2 years Crawford Station - Plant Eng. and then Mech. Maintenance Form. 1967 - 1 1/4 year Saxton Nuclear - obtained NRC Operator License. 8 years TMI Unit 1 - Supervisor Operations 8/1/68, Plant Engineer 1/1/73, Unit Superintendent 8/1/74 to May 77, Obtained 3R0 License.

f. Interfaces

1. Offsite The Director of Technical Support will primarily inter-face with the GPUSC of f-site technical support group in areas requiring specialized engineering assistance which can't be eptformed by the on-site staf f. He will also interface with other engineering firms where design assistance is required. He will interface with Materials Management on procurement / contract problems. He will interface with Regal.atorf Bodies such as the NRC on technical areas involved with the Technical Specifica-tions.

2. TMI-2 His responsibilities are the same for TMI-2 as they are for TMI-1.

3.2.8 Shift Technical Engineer

a. Function -

The Shif t Technical Engineer will provide direct technical oversight of the plant reactor performance and associated safety systems in order to improve the safety of unit opera-tions and maintenance performance. 1457 I87 5-15 Am. 4

b. Responsibility The Shift Technical Engineer (NUREG 0579-Section 2.2.1.b) reports to the Director - Technical Support. He is respon-sible for providing on shif t engineering, technical and administrative support to the Operations staf f personnel. A Shift Technical Engineer is on site at all times and can be in the Control Room within a few minutes,

c. Authority The Shift Technical Engineer acts in a monitoring / advisory capacity to the Operations Shif t Supervisor and Shif t Foreman.

He will advise and assist the Shif t Supervisor in matters of reactor safety by providing direct technical assistance or advising the Shift Supervisor whenever he believes it neces-sary to call for outside technical support. He has nn direct authority.

d. Minimum Qualifications The Shif t Technical Engineer shall have a Bachelor of Science Degree in an Engineering or Scientific related field. A minimum of two eyars of related experience in power genera-tion. In addition to the academic education, the Snift Engineer shall possess a thorough knowledge of plant systems and components. In addition, it is intended that the Shift Technical Engineer obtain the training necessary to be licensed as an SRO or on as soon as practicable basis but need not be licensed.

e. Incumbent Qualifications Incumbent A High School Graduate. University of Missouri - B.S. Mechani-cal Engineering - 1972.

1972-76 Field Engineer, General Electric Co. Installation & Se rv ice . Div . 1976-Present - Metropolitan Edison Company 1976 - Eng. II Nuclear TMI 1978 - Eng. III - Generation - Reading 1979 - Shif t Engineer III - TMI 6/66 to 4/70 U.S. Air Force - Inventory Management Specialist , Dyess Air Force Base , Abilene, Texas. Inbumbent B High School Graduate. North Carolina State University - B.S. Nuclear Engineering - 1976. 4/70 to 4/76 National Guard special schools in Accounting and Radar techniques. 1457 188 5-16 Am. 4

1970-1971 - HP Tech. with Westinghouse Nuclear Fuel Div. 1971-1973 - Service Representative 3M Corp. - Instrumentation Technician. 1976 - Present - Metropolitan Edison Company - Engineer I-Generation. 1979 - Engineer II - Generation 1979 - Shif t Engineer II - TMI. Incumbent C High School Graduate. 1976 - B.S. Physics Albright College and MSE Towne School of Engineering and Applied Science, University of PA. 1976 - Metropolitan Edison Company - Engineer I - Generation. 1979 - Metropolitan Edison Company - Engineer II - Generation. Incumbent D Righ School Graduate. B.S. Nuclear Engineering - Rensselaer Polytechnic Institute, Troy, NY - 1976. Master of Engineering - Nuclear Engineering - Penn State University 1979. 6/78 to 9/79 - Metropolitan Edison Company. 6/78 - Engineer I - Generation 9/79 - Shif t Engineer I - TMI Incumbent E Righ School Graduate. BSEE, Penn State University 1977. U.S. Navy 7/68 to 7/72 - Aviation Electricians Mate-Class A. Fligh Electrician and ECM Operator. 11/77 - Present - Metropolitan Edison Company - TMI. 11/77 - Eng. I - Nuclear 8/79 - Eng. II - Nuclear 9/79 - Shift Engineering Incumbent F Righ School Graduate. B.S. Nuclear Engineering - Penn State University 1977. 6/77 - Present - Metropolitan Edison Company - TMI. 6/77 Eng. I - TM I 9/79 Shif t Technical Eng. - TMI.

f. Interfaces

1. Offsite The Shif t Technical Engineer will not normally have any offsite interface. He may , howev e r , a t times seek assistance from support organizations such as B&W, GPUSC, etc. and will assist in the overall interface required during an emergency with the various outside agencies.

1457 189 5-17 Am. 4

2. TM I- 2 The Shift Technial Engineer at present is mainly concerned with Unit 1 but is also monitoring the Unit 2 Support Systems. Upon ar.signment of Unit 2 Shift Technical Engineers there should be little or no interface with Unit 2.

5.2.9 Manager of Support Services and Logistics

a. Function The Manager of Support Services and Logistics reports to the Vice President - Nuclear Operations and is responsible to plan, organize, and direct the day to day activities of the Personnel, Budgets and Reports, Security, Facilities, Docu-ment Control, Of fice Management , and Safety functions.

b. Responsibilities This position assures that ef fective programs are implemented in Personnel Administration, Personnel Recruiting and Employ-ment, Personnel Wages and Salary Administration; that an effective Operations and Maintenance and Capital Budget preparation, review and approval process is in ef fect, that the TMI Security plan is implemented in accordance with all applicable regulations; that an efficient Office Management program is in effect; and that the Industrial Safety Program is developed , organized , and implemented to insure that a safe working atmosphere exists for all employees and that all applicable safety regulations are met,

c. Authorities Consistent with the responsibilities of this posirion, the Manager of Support Services and Logistics is authorized to institute procedures required to implement programs which improve or enhance the degree of Support Services provided to the plant operating staff. Additionally, consistent with the level of signature authority established, this Manager can extend job offers and approve other personnel transactions, authorize and approve expenditures and authorize facility changes.

d. Minimum Qualifications This positicn is rr quired to support the Met-Ed organization and shall have appropriate qualifications.

5-18 145/ 190 Am. 4

e. Incumbent Qualifications The (acting) incumbent received a Bachelor of Science Degree in Civil Engineering in 1957 f rom the Pennsylvania State Univ e rsity. From 1957 to 1959 eh served as a Navy Of ficer on an Amphibious Ship. He was employed by Met-Ed in 1959 and has held positions in the Tr nsmission Engineering function for fourteen years, the c. tribution Operations function for one year, the Operations Analysis function for four years, and is currently Manager - Generation Administration, having been named to that position on April 1,1978. He has attend-ed the Public Utility Executive Program at the Graduate School of Business Administration, University of Michigan,

f. Interfaces

1. Off-site A functional line of communication exists with the GPUSC Manager - Management Services to coordinate intercompany actions in the functional areas of responsibilities assigned to this position.

2. TMI-2 All functional areas of responsibility assigned to this position are common between TMI Unit 1 and TMI Unit 2; therefore, the employees in this organization unit routinely divide their time evenly between both units.

5.2.10 Superintendent - Radiological Controls and Chemistry

a. Function The Superintendent - Radiological Controls and Chemistry reports to the Vice President - Nuclear Operations on all matters affecting the Radiation Protection and Health Physics aspects directly associated with the operation and mainte-nance of Unit I plus the technical review and guidance regarding the Chemistry Program of Unit I.

b. Responsibility The Superintendent - Radiological Controls and Chemistry is responsible for:

The development and operation of a Radiation Protection Program which meets or exceeds that specified in the Radia-tion Protection Plan, Technical Specifications, Federal Regulations or other Regulatory directives. 1457 19i 5-19 Am. 4

The coordination of the Radiation Protection Department with the activities of other Unit I Depc rtments. The direction of the ALARA Program including the technical review of all matters addressed to that Department which involves personnel exposure to the hazards of external of internal radiation. The technical review of all operations which release or could possibly release radioactivity to the environs. The conduct and documentation of monitoring of all radioac-tive material received or tran ferred by Unit I. The maintenance of records reflecting the results of all inspections and surveys pertinent to the Radiation Protection Program of Unit I. The technical review and approval of all courses of training which may affect the Radiation Protection Program including the approval of all examinations which measure the effective-ness of such training. The certification of qualification of all personnel who perform radiation monitoring or survey functions directly affecting the radiological safety of other personnel. The support of the Unit I Superintendent through the technical review of the chemical and/or radio-chemical analyses and an independent monitoring of the performance of the conduct of such analyses. The reporting to proper station authority all unusual operat-ing conditions likely to affect personnel safety or plant reliability.

c. Authority The Superintendent - Radiological Controls and Chemistry has the authority to unilaterally direct the termination of any operation or activity which may adversely affect the radio-logical health or safety personnel or the environment.

d. Minimum Qualifications The Superintendent - Radiological Controls and Chemistry shall possess the minimum qualifications set forth in Regulatory Guide 1.8. Specifically, the ..cumbent should have a Bachelor's Degree or the equivalent in a Science or Engineering subject including some formal training in radia-tion protection. The incumbent should have at least 5 years' 1457 192 5-20 Am. 4

of professional experience in applied radiation protection. At least three years of this professional experience should be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encoun-tered in a nuclear power station, preferably in an actual nuclear power station.

e. Incumbent Qualifications High School Graduate. Villanova University - BME Degree.

U.S. Navy 6/54 to 8/78 Retired. Rank: Captain 1 year experience Nuclear Power Plant - TMI (Technical Administration Support).

f. Interfaces

1. TMI-2 A separate organization reporting to the Director - Unit 2 Recovery, in addition to providing similar services to Unit 2, provides those common station functions which include personnel dosimetry, respiratory protection technical support , maintenance and calibration of all survey instrumentation, and the shipment of all solid radioactive waste.

5.2.11 Manager Training

a. Function The Manager - Training reports to the Vice President-Nuclear Operations. In this position he is responsible for the operator training, technician training, and accelerated operator retraining. The technical training section will include training for maintenance and health physics techni-cians. The oeprator accelerated re-training program is a broad program based upon changes and lessons learned as a result of the TMI-2 accident. The training department will be augmented by outside consultants as necessary.

b. Responsibility This position is responsible for the training of all personnel at TMI. This responsibility is discharged through an organi-zational structure devised to address the trainee along functional discipline lines. In addition, the organization provides its own administrative support and any temporary personnel necessary to conduct of major programs. Principle acocuntabilities of this position are:

1457 193 5-21 Am. 4

1. To assure the content and conduct of training for Reactor Operators, Health Physics and Maintenance Technicians, Professional personnel and Management personnel meet Federal, State, and Local regulatory requirements and conform to industry standards.

2. Oversee the development and maintenance of a viable station training philosphy.

3. To select, develop and mold a highly competent training staff.

4. To nurture and encourage a positive training attitude among the trainees.

c. Authority The incumbent Manager - Training possess expertise useful for providing operational advice and guidance to plant staf f personnel in the event of unusual occurrences. This exper-tise will be utilized in event of such occurrences.

d. Minimum Qualifications This position is required to support the Met-Ed organization and shall have appropriate qualifications.

e. Incumbent Qualifications B.S. Degree in Physics - Ill. Institute of Technology Employed - U.S. Navy 1948 Reactor Operator, Nuclear Power School Instructor.

Met-Ed - 1973-Present - Head of Licensing, QA Program, Manager Generation Operations. Argonne National Laboratory 6/58-7/68 - Instructur and Reactor Operator. Vermont Yankee Nuclear Power Corp. 7/68-7/73

f. Interfaces

1. Offsite The position, Manager - Training, interfaces with Licens-ing primarily in the functions of interpretation regula-tory requirements for training and LER review. Interface with QA consists of discussion and resolution of training related audit findings. Interface with Eng. is for the purpose of utilization of knowledgeable and qualified engineers to develop training materials and provide instruction.

1457 194 5-22 Am. 4

2. TMI-2 Insofaras TMI-2 operations personnel participate in applicable Unit 1 traning programs and no additicnal training department resources are required for thel-training. Uait 2 Ops. Dept, training on unique materisl utilizes app - Aim %tely three of 13 operator training section personnel. Training for all other TMI-2 person-nel is the responsibility of a separate group.

5.2.12 Superintendent of Maintenance

a. Function This position reports to the Director of Maintenance shown on figure 5.3-1. The Superintendent of Maintenance is part of the offsite organization but is physically located on site and works closely with the onsite staf f. Reporting to this position are two (2) Supervisors Unit Maintenance - Nuclear, Utility Supervisor, Technical Analyst Senior - Nuclear, Parts Coordinator, Administrator Maintenance - Nuclear and Welding Foreman.

b. Responsibility This position is responsible for planning , organizing, integrating and directing the corrective maintenance effort of the TMI nuclear generating station in order to insure optimum equipment / systems availability and reliability.

This position is responsible for assuring the provision of adequate resources to carry out the Maintenance Programs.

c. Authority Resolves resource allocation problems. Commits company and contractor resources.

d. Minimum Qual.ifications This position shall have seven (7) years of responsible power plant experience or applicable industrial experience, a minimum of one (1) year which shall be nuclear power plant ex pe rience. An Associate Degree in an Engineering or Scien-tific field is preferred and may be credited to the remaining six (6) years of experience. The individual should have non-destructive testing familiarity, craf t knowledge, and an understanding of electrical, pressure vessel and piping codes.

1457 195 5-23 Am. 4

e. Incumbent Qualifications The Superintendent of Maintenance is a Navy veteran of twenty-seven years with experience in Marine and Power Plaat Operations, Maintenance and Repair. He has served as Chief Engineer and Repair Of ficer on several large combatant Naval ships. He was a member of the Naval Board of Inspection and Survey as an engineering inspector. He attended engineering service schools as a prerequisite for qualification as Engineer Officer. He was appointed Supervisor of Maintenance at TMI in 1973.

f. Interfaces

1. Offsite Contractors, vendors representatives.

2. TMI-2 Supervisor - Maintenance - Unit 2.

Waste Management Manageer. 5.2.13 Supervisor of Maintenance

a. Function This position reports to the Superintendent of Maintenance-Naclear as shown on Figure 5.3-1 (also see 5.2.12). Re po rt-ing to this position are the Lead Mechanical, Electrical and Instrument and Control Foremen, Jr. Planner, and six (6)

Shift Maintenance Foreman,

b. Responsibility This position is responsible for planning, organizaing, integrating and directing the maintenance effort for a generating unit of the TMI nuclear generating station in order to insure optimum equipment / systems availability and reliability.

This position is responsible for the identification, justifi-cation and utilization of resources requisite to the mainte-nance program.

c. Authority Commits company manpower and materials. Authorizes procurement /

utilization of materials and supplies. Commits Contractor resources. 1457 196 5-24 Am. 4

d. Minimum Qualifications This position is required to have seven (7) years of respon-sible power plant experience or applicable industrial ex-pe rience , a minimum of one (1) yea which shall be nuclear power plant experience. An Associate Degree in an Engineerind or Scientific field is preferred and may be credited to the remaining six (6) years of experience. The individual should have non-destructive testing familiarity, craf t knowledge, and an understanding of electrical, pressure vessel, and piping codes,

e. Incumbent Qualifications The Acting Supervisor of Maintenance received a B.S.M.E. in 1968 from the University of Missiouri.

1968-1979 U.S. Navy, Naval Nuclear Power School - Vallejo, CA, Naval Nuclear Power Training Unit - Idaho , Naval Sub-marine School. Nuclear Attack Submarine - USS Jack; Engineering Officer of the Watch, Communications Of ficer, Electrical Of ficer, Main Propulsion Assistant, Qualified: (1) in submarines, (2) as Nuclear Engineer Officer; Ship - 18 month non-refueling overhall Portsmouth Naval Shipyard. Nuclear Missile Submarine - USS John Marshall; Navigator / Cperations Of ficer (3rd Of ficer), Engineer Of ficer, Executive Of ficer (2nd in Command); Ship - 22 month refuleing overhall in Mare Island naval shipyard. Separated from the Navy as a Lieutenant Commander. TMI - Special Maintenance Projects - 6 months,

f. Interfaces

1. Offsite Contractors, vendors' representativ es .

2. Other TMI Staff QC Supervisor, Radiation Protection Supervisor, PM Supervisor, Director Technical Functions , Unit Superin-tendent Operations Supervisor.

3. IMI-2 None 1457 197 5-25 Am. 4

5.2.14 control Room Operator

a. Function Each Control Room Operator reports to the Shif t Foreman and oeprates the reactor, turbine, generator, switchboards and all other equipment necessary to maintain continuous produc-tion with maximum safety and efficiency in accordance with the Operating License.

b. Responsibility The Contral Room Operator is responsible for all the equip-ment assigned to him in the Unit, and the reporting of any unusual performance of this equipment to the Shift Foreman.

in this regard the Control Room Operator:

1. Plans, directs and coordinates the work of others working with or assisting him and cooperates with other Control Room Operators on the same shift.

2. Is responsible for the oepration of all equipment assigned to him in the power station and reporting of any unusual performance of this equipment to the Shift Foreman. Makes recommendations to improve the design, operation or performance of the station.

3. Operates or controls nuclear reactors, turbine generators, switchbcards and nll other equipment to maintain continu-ous required production with maximum safety and ef ficiency in accordance with the Operating License, Technical Specifications, normal and special operating procedures.

4. Observes and interprets indications from detailed and complex instrumentation systems.

5. Checks all orders received from the Dispatcher and is responsible for their completion or for informing the Shift Foreman of all orders that cannot be completed; synchronizes generators, lines or other electrical equipment as required; distributes load on generating equipment; maintains proper voltage; calculates and records designated information on logs and reports.

6. Operates switches accurately according to own judgment or predetermined regulation.

Is responsible for issuing all orders for switching and tagging of all electrical equipment of the station not under the jurisdiction of a Dispatcher. Is responsible for issuing all orders for mechanical tagging of all mechanical equipment of the station. 1457 198 5-26 Am. 5

Is responsible for administration of the Company's tagging and safety rules insofar as they relate to switching and tagging orders.

7. May assist supervisors in training of other personnel in the operation and control of all equipment for which he is responsible.

8. Performs other related or lesser skilled duties for which he is qualified or has received proper instruction or direction.

c. Authority The Control Room Ojerator has the authority to shutdown the unit when condition 0 in the unit warrant such action. He also has the authority to direct the auxiliary operators in their performance of Company approved procedures.

       '. Minimum Qualifications High School diploma or equivalent and two years experience ina power plant one ow which is at a nuclear plant. The Control Room Operator must be licensed by the NRC.

5.2.15 Auxiliary Operator -

a. Function The main function of each Auxiliary Operator is to operate and inspect equipment in the nuclear power station as re-quired to support day-to-day operaton from his position outside the control roo. He will be directed in the per-formance of his duties by the Shif t Foreman or the Control Room Operator and will report any unusual performance of equipment to either one of them.

b. Responsibility The Auxiliary Operator shall be responsible for the operation and inspection of plant equipment. He also performs the functions of radiation protection monitor on his shif t as required and is responsible for notifying the appropriate supervisor if any portion of the unit exceeds established limits. He will assist in the receipt, storage, loading and unloading of furel, shipment of irradiated materials and disposal of racioactive wastes as directec.

c. Authority The Auxiliary Operator shall have the authority to exec.ute Company approved procedures as directed by control room operators or shift foreman.

1457 199 5-27 Am. 5

d. Minimum Qualifications Each Auxiliary Operator should have a high school diploma or equiv alent . The auxilairy operator shall be interviewed to verify that they exhibit mature judgment, testing willb e used to aid in determining the individuals ability to progess to higher levels of responsibility and eventual NRC licensing.

5.2.16 Technical Analyst Sr. I (Fire Protection)

a. Function This position coordinates the Fire Protection Program at Three Mile Island.

b. Responsibility This position is responsible for providing engineering, mnagement, maintenance / test scheduling, and outside vendor and inspection group coordination for the TMI Fire Protection Systems. This includes procedure preparation and review, management and coordination of survellance and special fire systems tests, arranging for outside contractor support, parts review, development of specific fire fighting plans and significant contact with NRC and NELPHIA inspectors and safety and Generation Engineering personnel.

c. Authority The Auxiliary Operator shall have the authority to execute Company approved procedures as directed by control room operators and shif t foreman.

d. Minimum Qualifications Righ School Graduate or equivalent and appropriate related expe rience.

e. Incumbent Qualifications Righ School Graduate 1966 U. S. Nav y 1968-1977 Nuclear Power School Engineering Lab. Tech. and Casualty Control.

f. Interface The position also interfaces with the company Fire Protection Engineer (Director of Fossil Plants) and coordinates the fire protection activities for both units.

1457 200 5-28 Am. 5 l

5.3 STATION SUPPORT ORGANIZATION The facility organization is supplemented by the resources of General Public Utilities. The GPU Station Support Organization, shown in Fugure 5.3-1, will function in the five main areas of: corrective maintenance , D41-2 recovery, technical f unctions, environmental health and safety and reliability engineering. Th" Director of Maintenance reports to the Senior Vice President-Mr.c-Ed-Vice President - GPU and is responsible for corrective v.41ntenance and construction at Three Mile Island including coordination with outside contractors as necessary. The Superintendent of Maintenance reports to the Director of Maintenance and is responsible for Met-Ed and contractor cor-rective maintenance on TMI-1 and 2. Operations related cor-rective maintenance activities are coordinated through the respective supervisors of maintenance with to< unit supervisors of operations. A Supervisor of Maintenance on each unit reports to the Superin-tendent of Maintenance and is responsible for Met-Ed corrective maintenance in the mechanical, electrical, and instrument and control areas for the respective units. Lead foreman on each unit in the various maintenance areas report to the supervisors of maintenance as shown in Figure 5.3-1. The Director of TMI-2 Recovery reports to the Senior Vice Presi-dent Met-Ed-Vice President GPU and is responsible for the licens-ing and environmental safety of generating stations. The Director - Environmental Health and Safety reports to the Senior Vice President Met-Ed-Vice President GPU and is respon-sible for the licensing and environmental safety of generating stations. The Director Technical Functions reports to the Senior Vice President-Met-Ed, Vice President GPU. In this position he will be responsible to provide a centralized technical capablity to support generating facilities. This capability will include general mechanical, civil, electrical and instrumentation and engineering mechanics areas to assist in the solution of plant operating problems. In addition, this position will be respon-sible for supporting GPU nuclear plants in the areas of nuclear fuel management, process computer, control and safety analysis, and plant operational analysis. In addition, TMI Engineering Management section has been organized to be the focal point for the coordination of all out of plant technical support for TMI ope ra tions . 1457 201 5-29 Am. 5

The Director of Reliability Engineering reports to the Senior Vice Preside t-Met-Ed Vice President GPU. In this position, he is responsible for the activities of the systems laboratories and the quality assurance functions including the quality control support on site. 5.3.1 GPUSC Technical Functions Group The GPUSC Technical Functions Group consists of four engineering departments, namely Systems Engineering, Engineering & Design, Technical Management and TMI-2 Recovery Engineering. This group provides offsite support for TM1-1 as specified below: All staf f, except as noted, includes only GPUSC permanent person-nel. Support from outside contractors is not included but is available on short notice to supplement the GPUSC staff as necessary. This may be used to accommodate short term manpower intensive needs or accommodate temporary vacancies. The Technical Functions Group management personnel each have at least a B.S. degree in engineering or science and the following experience. Years of Eng. Years of Nuc. Title Experience Experience Director, Technical Functions 26 24 Manager, Systems Engineering 22 22 Manager, Engineering & Design 20 14 Manager, TMI-2 Recovery Engineering 14 14

a. Systems Engineering Department is responsible for providing support in the areas of nuclear fuel management, process computer, control and safety analysis and plant operational analysis. More detailed responsibilities are:

1. Nuclear Analysis & Fuels - Responsible for analytical and other activities related to core reloads, fuel management and the physics performance evaluation, including shield-ing analysis.

2. Process Computers - Responsible for all process computer systems including both computer hardware (main frame and auxiliary memory units, input / output equipment and CRT display devices) and computer sof tware.

3. Control & Safety Analysis - Responsible for plant control systems engineering. Plant subsystcm dynamic simulation and safety analysis.

4 Plant Analysis - Responsible for analyzing overall plant performance and the behavior of individual systems and Components. 1457 2.02 5-30 Am. 5

The current department staff is 30, with a total staff of 51 planned (includes 7 Met-Ed personnel for ?racess Computers).

b. The Engineering & Design Department provides a centralized capability in the general mechanical, civil, electrical and instrumentation, and engineering mechanics areas. Other services include civil engineering and engineering standares and procedures. More detailed responsibilities are:

1. Engineering Mechanics - Technical expertise for the analysis of all structural and fluid mechanics problems, including piping, stress analysis and supports, general vibration and dynamics of mechanical equipment, acoustic noist, and fluid dynamics such as water and steam hammer, cavJtation and related problems.

2. Mechanical Systems - Primary responsibilities for the analysis, engineering and design of all fluid and ma-terials handling systems.

3. Mechanical Components - Provides technical expertise in the application of specialized mechanical components including pumps, fans, valves, heat exchangers and power conversion equipment. Also responsible for the general area of water treatment and industrial waste management systems.

4. Electrical Power & Instrumentation - Responsible for analysis and design of plant main and auxiliary electrical power distribution systems, protective relaying, light-Ing, communications, and grounding and cathodic protec-tion and design capability for instrumentation application and I&C circuits.

5. Design & Draf ting - Responsible for providing generalized design and drafting support, including piping systems, electrical power and instrumentation, plant arrangements and equipment installation.

The current department staff is 53; with a total staff of 74 planned.

c. The Technical Management Group is responsible for coordinat-ing and directing of assigned projects including all out of plant technical support for TMI operations.

The current group staff is 8 with a total staff of 12 planned.

d. The TMI-2 Recovery Engineering Department is responsible for coordinating and directing assigned projects dealing with IMI-2 recovery, including radwaste reactor systems and containment and in-containment restoration projects. This depa r tment also provides onsite engineering support in the following areas:

57 203 5-31 Am. 5

Fluid and electrical systems, radiation and radiochemistry analysis, and process and waste management systems. The current department staff is 4 with a total staff of 20 planned. 1457 204 5-32 Am. 5

Director of Reliability Engineering. The Manager of Quality Assurance and the Director of Reliability Engineering are inde-pendent of de sign, procurement, manuf acturing , construction, operations, or maintenance and report at a suf ficiently high enough level to provide an independent assessment and evaluation of the ef fectiveness of the implementation of the QA Program. The Manager of Quality Assurance has the overall authority and organizational freedom to identify quality or management control problems and provide recommended solutions. This authority and responsibility includes the stoppage of work or the recommenda-tion that an operating nuclear unit be shut down, the Manager of Quality Assurance has direct reporting authority to the Senior VP GPUSC/VP Met-Ed and shall use this path when difference ci opinion with the Generation Group regarding quality arise. The re-organized Quality Assurance Department consists of five major sections as shown in Figure 5.4-2. Listed below is a descriptiono f the responsibilities of each section together with a discussion of the major differences from the pre-accident Metropolitan Edison - Operational Quality Assurance Department. Changes depicted result in part from the analysis following the TMI-2 accident: (1) Design and Procurement Assurance Section - This section with a quality engineering staff located both in the corporate head-quarters and at TMI constitute the main technical support sections establishing quality programs, and inspection requirements in support of design and procurement ac tiv itie s . The same gorup reviews quality related materials and product specifications and procurement requisitions to assure that the committed to require-ments have been established. Additionally, this group is involved with evaluation of specific vendors (contractors) and their programmatic controls against established requirements. An element of the on site Design and Procurement Assurance Section has the responsibility for reporting quality trending and performing final verification and acceptance of installation / modification documentation packages before turnover to Records Storage. The major change, organizational 1y, from the previous Quality Assurance Organization is the formation of the on-site quality assurance / design element to input quality related information and perform reviews in support of design and procurement. (2) Manufacturing Assurance Section - The prime responsibilities of this section are to perform those necessary post award quality related activities requireo to assure that vendor's product is designed, manufactured, and tested in accordance with those specified quality requirements. Trend information supplied by this group weighs heavily in the maintenance of the Vendor's Clasd fication List. 1457 205 5-3' Am. 4

The major change organizationally is that this is now a separate organization section. (3) Modifications / Operations Section - This section consists of two major sub-groups, Quality Control and Operation Quality Assurance. Quality Control is responsible for receiving in;eection and the inspection and/or surveillance activities related to corrective maintenance, modifications, installation or new construction. The group has specialists who are qualified to the appropriate levels of ANSI N45.2.6 and SNT-TC-1 A. Additionally, the group has a welding engineering section which reviews contractors procedures and surveils control of special processes. Operational Quality Assurance is responsible for monitoring functional testing and performing surveillance of all operations ac tiv itie s . The latter includes monitoring and surveillance of plant operations, preventative maintenance, radiation protection and the processing, packaging and shipping of contaminated products , and radioactive wastes. The Operational Quality Assurance group is also responsible for in-service inspection and monitoring performance and results of pump and valve testing to the applicable requirements of ASME Section XI. The major difference from the previous organization is in the formation of the separate operational QA group for surveilling compliance with the technical specification requirements. The other revision is the inclusion of in-service inspection require-ments under QA Department scope. (4) Methods, Operations and Audit Section - This section is responsible for QA Department program development. It is, therefore, responsible for coordinating activities associated with department procedures and indoctrination and training. Additionally, the group conducts independent evaluation and assessment of the program's implementation thru Quality Assurance Audit Program. The latter includes an evaluation of ef fectiveness of the pro-grammatic aspects of the QA Program. This program satisfied the requirements of ANSI N45.2.12 and utilizes auditors qualified to ANSI N45.2.23. Assisting in this assessment is a full-time site audit group reporting independently to the Manager of Quality Assurance and the Director-Engineering Reliability thus providing management assessment of the ef fectiveness of the program. Additionally , both ections are available to provide timely close out and verification of identified problems. 14^$7' 206

(5) Materials Technology Section - This is an off-site section which has the responsibility of supporting design in establish - ment and/or review of requirements. Additionally the group is available as a staff group to support Manufacturing, Construction and Operations in assessment and/or evaluation of identified materials technology problems. To help af fect the implementation of thi responsibility are the services of the of f-site metallur-gical ..boratory which now reports to the Director-Reliability Engineering. The specific services provided by the Materials Technology section include: Non-destructive Examination In-Service Inspection Materials Engineering Welding Engineering Whereas, other sections have full-time technical expertise in these areas, this centralized gorup will provide technical direction. The Materials Technology Group did not exist in the Met-Ed Organization although elements of materials technology function existed throughout the Met-Ed/GPUSC Organizations. 5.4.3 Program The QA Program is in the process of undergoing major revisions and as previously identified to the NRC on October 8,1979 (GQL-1233), this revision is scheduled for submittal on January 15, 1980. The revision was necessitated for the following reasons:

a. To describe the newly formed TMI Generation Group Organiza-tion and identify the management controls and Quality Assurance Program for this Organization. Many of the or-ganizational changes depicted are reflections of careful analysis resulting from the TMI-2 accident and our desire to provide a more effective and responsible management control over operations having an affect on nuclear safety.

b. To provide a workable QA program plan that describes the functional manner in which activities af fecting safety are managed. Included but not limited to those activities are start-up, shutdown, normal operation, emergency actions, surveillance, maintenance, repairs, radwaste processing, and modifications.

5-35 Am. 4 1457 207

The TMI Generation Group is committed to a comprehensive quality assurance program consisting of a three level approach to assure satisfactory and complete implementation of the program commen-surate with its requirements for safety and performance. The program's foremost considerations are the protection of the general public's health and safety. The three level approach is defined below: Level 1 - activities at this level include independent inspections, checks and tests. This level of activity may be performed by the Operations Department such as surveillance tests, calibration of instruments, radiation surveys , analy-ses of samples, etc., the Quality Control Section such as receipt inspection or inspections of modification or corree-tive maintenance activities, or by contractors as part of their scope of work. In all cases, the activity is performe<.. by individuals knowledgeable of the activity being performed and qualified to perfc rm *he w;ck. Checklists or data sheets are also used for documenting the results of activity and for providing a permanent clant record of the performance of the activity. Level II - The activities at this level are primarily those of surveillance or monitoring and are performed as deemed necessary by the Operational QA, Quality Control, or Manufac-turing Assurance departments. The level of surveillance / monitoring applied is consistent with the importance of the item to safety. For activities, whereby QC is performing first level inspection, no second level activity will be required. At this level procedures and instructions are established and surveillance records will be completed and maintained. Such surveillance / monitoring normally includes observation of quality control tests and inspections, observation of signi-ficant operations, review of records, verifications of test reports, and direct over inspection on.a spot check basis. The organizations performing this activity have these levels of authority, the lines of internal and external communica-tion for management direction and the properly trained personnel for implementation of these activities. Level III - The purpose of this level of activity is to assure chrough a comprehensive program of review and auditing that the first and second levels of the program are properly functioning. The purpose of the program is also to establish that all other organizations including Operations, Mainte-nance, Engineeting, Materials Management,e tc. are properly satisfying all the requirements of the Operational QA Program. 5-36 Am. 4 1457 208

At this level procedures and instructions are established including the use of comprehensive checklists for documen-tation of the audit or third level activity. The program requirements of ANSI N45.2.12 are satisfied. Qualified audit personnel are included that satisfy the requirements of ANSI N45.2.23. Additional technical experts from areas with administrative reporting outside the function that is being audited will be included as the Audit Team Leader deems necessary. The organization performing this activity has sufficient authority and lines of internal and external communica* ions for obtaining the nacessary management direc-tion. In addition to the Quality Assurance Department's actiities associated with verification of the completeness and adequacy of the work performed there also are several independent review groups whose responsibilities are to provide independent safety review and operational advice. These groups each have che technical expertise in the areas of licensing, operations, and radiation protection. The groups include the following: General Of fice Review Board (GORB) - This group is an advisory group to the President of GPU and whose responsibilities are to foresee potentially significant nuclear and safety problems and to recommend to the President how they may be avoided. Generation Committee (GRC) - This group is assigned the responsibility to provide independent review and audit of designated areas of operations, nuclear engineering, chemis-try and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering and quality assurance practices. The GRC reviews safety evaluations, significant operating abnormalities, violations of codes, Tech Specs, etc. and deficiencies. Plant Operations Rc<iew Committee (PORC) - This committee functions as an advisory group to the Unit Superintendent on all matters related to nuclear safety. Radwaste Review Committee (RRC) - This committee functions to provide independent review and audit of designated activities in the areas of radioactive waste management, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering and quality assurance practices associated with radioactive waste materials ccatrol and management. The scope of the quality program while defined in detail in TMI procedures in general includes crierials, components, systems, and structures identified by Engineering as falling uner the following areas of concern: 5-37 Am. 4 1457 209

a. Nuclear safety related

b. Fire safety related

c. Rad / Waste as required by RG 1.143

d. Rad / Waste shipments as required bv 10CFR70 Appendix E

e. Quality Control augmented The extent to which the program is applied is detailed in the Operational QA Program.

The Operational QA Program includes provisions for controlling, and identifying the in process, final inspection, examination, test and oeprating status of all equipment, structures, compo-nents and systems throughout the plant. This control will include provisions which preclude the final use of structures, systems, and components which are non-conforming or removed from service for purposes of maintenance, modification, inspection or test. 5.5 STATION ORGANIZATION UNDER ACCIDENT CONDITIONS The TMI-1 Emergency Plan (See Section 4.5.1.3) incorporates all the current requirements in the area of emergency planning. It is anticipatid that it will be revised once comments from the NRC Task Force ot Emergency Planning are received. The Emergency Plan is to be implemcated 60 days prior to the restart of TMI-1. Upon final pla . approval, specific personnel assignments will be made to f11' the emergency organization positions. It should be noted that the Emergency Plan provides for manning the on site emergency organization within one hour and the off site emergency organization in a two to four hour time frame. 1457 210 5-38 Am. 4

Senior Vice-President Met-Ed Vi ce-P r et i <t e n t CPOSC _ vic.-Presider.t Nuclear Operations - __ birect3r Technical Support h ui.i t i k Supe r in t endent Manager Superintendent Manager Support Servtece Training Radiology Controls and Logistics and Chemistry Lead ::uclear E Operator Security Engineer Supervisor Supervisor Training Preventative Operations Section Main t e nance Unit I - Rad Waste Facilities. Lead Electrical k Unit 1 3 Office Nnagement, E n t. ir.ce r Personnel Tec h.4 t c ian l Lead !&C d Llectrical Shif t

• T ra in ir.g Wespiratory

       ]       Fv.Inver                                      Su pe rv i so r            Section               Protection W                                   -   Radiological Control Engineer (ALARA)

Dosimetry Operate.r Leal i:echanical 5 &:c han ica l Shift " Accelerated Enginetr Foreman Retraining ' Chemistry dI&C j @ Program Shift Technical k Engineec* Control Room Opera tor * , Fire Protection b Auxiliary @@ Operator

• LEGEND Formal Education z.

LT1 y f License Required C

   .if                                                                       Figure 5.2-1          Station Organization              y Y

N o *Six (6) Shift Operations f e

i Senior Vice-President Met-Ed Vice-President GPUSC Director Director TML-il U1 rector otrector Ma inte nance Recovery Technical -Environmental Reliability Functions realth & Safets Engineering Maste Ma nagement Engineering Licensing Quality f _ Su pe rint endent Manager De sign Assurance Maintenance Construction Project - Environmental Management ~ Systems Systems Su pe rvi so r Su pe rvi so r Cont r ac to r Construction Engineering La bs Maintenance Ma i n t e na nce Corrective Activities U,1t t , Unit II Maintenance ll PHI-Il Recovery Mechanical Mechanical THI Engineering Co r rec t ive Co rrec t ive stinagement Main t e na nce Ma in te na nce Un n i Unit II Electri-al Electrical Corrective Cor rec t iv e Maintenance Maintenance Unit I Unit 11 ___. Figure 5.3-1 Instrument Instrument Station Support Organization

,g3, and Control          and Control k Il _ Co r rec t ive       Corrective N'l    Ma in te na nce      Maintenance Unit I               Unit II N

TMI GENERATION GROUP Senior V.P. V.P. Nuclear Operations Director Director Director Director Director Environment, Maintenance Technical TMI-2 Recovery Reliability Health & Safety Functions Engineering FIGURE 5.4-1 1457 213 Am. 4

RELIABILITY ENGINEERING

              ,UALITY Q       ASSURANCE DEPARTMENT Director of Reliability Engineering Manager of Quality Assurance i

Methods, Operations Materials Technology and Audit Section Section Design and Manufacturing Modifications / Procurement Assurance Operations Assurance Section Section Section FIGURE 5.4-2 1457 214 Am. 3

1 u, A N m 0 a Z v' 1457 215

6.0 OPERATOR ACCELERATED RETRAINING PROGRAM (OARP)

6.1 INTRODUCTION

In preparation for restarting IMI-1, a retrainin6 program for TMI-l Reactor Operators and Senior Reactor Operators is being implemented. Several training issues considered as prerequisites to resuming power operation at TMI-1 have been identified and addressed in the Operator Accelerated Retraining Program and subsequent evaluation process is required of all personnel who will be assigned as Reactor Operators and Senior Reactor Opera-tors at TMI-1 during the resumption of power operation. The Operator Accelerated Retraining Program includes over sixty (60) presentations and/or practice sessions involving over two-hundred hours of training. Included in the program are at least twenth (20) hours of training directly involved with analyzing and handling abnormal and emergency situations at the Babcock and Wilcox Nuclear Training Center Simulator. The Operator Accelerated Retraining Program covers topics which can be grouped into four f unctional areas:

• TMI Plant System Review
• TMI Plant Operational Review
• Radiosctive Materials Control
• TMI Plant Transient Analysis The combination of the Operator Accelerated Retraining Program and the previous TMI-1 operator training and requalification programs can enable the safe and ef fective operation of the Three Mile Island Nuclear Station Unit 1.

6.2 PROGRAM OBJECTIVES The Operator Accelerated Retraining Program is designed to accomplish several objectives relating to enhancing TMI-1 Reactor Operator and Senior Reactor Operator performance. The achieve-ment of these objectives is in accordance with the performance standards specified in Section VI (Evaluation Procedure) and is a prerequisite to resuming operation of TMI-1. Program lectures which support the objectives and references for the objectives are listed in Appendix A. The Operator Accelerated Retraining Program objectives are as follows: A. To improve operator performance during small break loss of coolant accidents. B. To assure that the operator can recognize and respond to conditions of nadequate core cooling. 1457 216 6-1 Am. 1 l

C. To improve operator performance during transients and acci-dents including events that cause or are worsened by inap-propriate operator action. D. To assure that the operators have an in-depth understanding of the TMI-2 accident and lessons learned. E. To assure that operators are knowledgeable of operating procedures and actions required upon initiation of the engineering safeguards features including reactor coolant pump requirements. F. To assure that operators understand the manometer effects of water levels in the reactor coolant system under dif ferent coolant system presure and temperature conditions. G. To assure that operators are aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains. H. To assure that operators are aware of the prompt NRC notifi-cations required in the case of serious events and signifi-cant ev ent s. I. To provide the operators with an in-depth understanding of the methods required to establish and maintain natural circulation. J. To assure that operators are knowledgeable of both short and long term plant systems modifications. K. To provide the operators with a review of the major plant systems. L. To provide specialized training on " Operations and Procedural Guidance Requirements". M. To assure operators are fully qualified through the 4.dmini-stration of the Company and NRC administered final wil." en. and oral examination. N. To provide the operator with a review of major administrative, normal, abnormal, and emergency procedures. O. To assure all licensed Unit 1 operators receive training on the B&W Simulator covering the TMI-2 incident. 6.3 Topical Outline The Operator Accelerated Retraining Program includes over sixty (60) presentations and/or practice sessions covering topics which can be grouped into four (4) functional areas:

• TMI Plant Systems Review
• TMI Plant Operational Review 1457 217 6-2 Am. 1 l
• Radioactive Materials Control
• TMI Plant Transient Analysis The program topics include coverage of essential information needed to understand TMI-1 plant design and oepration. Detailed information on plant systems, operating procedures, and transient analysis are also included to provide an overall understanding of safe nuclear plant operating practices.

A. TMI Plant Systems Review Topics which provide a specific plant systems information address the following areas:

• Features of Facility Design
• Instrumentation and Control
• Safety and Emergency Systems Presentations covering specific information on system func-tions, capabilities, limitation, interrelatiunships and controls are involved.

The specific topics are:

1. Reactor Coolant Syrram

2. Makeup and Purification System

3. Control Rod Drive System

4. Nuclear Instrumentation and In-Core Instrumentation

5. Decay Heat Removal

6. Decay Heat River System

7. Containment Isolation System

8. High Pressure Injection System

9. Nuclear Services Closed Cooling System

10. Decay Heat Closed Cooling System

11. Core Flood System

12. Nuclear Service River Water System

13. Reactor Building Emergency Cooling System

14. Intermediate Closed Cooling System

15. Feedater System

16. Condensate System

17. Emergency Feedwater System

18. Main Steam System

19. Electrical Distribution System

20. Emergency Diesel

21. Reactor Protection System

22. Ventilation

23. Hydrogen Recombiner and Hydrogen Purge

24. Emergency Safeguards Actuation System

25. Non-nuclear Instrumentation and Interlocks

26. Computer and Mod Comp

27. TMI-l Short Term Change Modifications

28. TMI-l Long Term Change Modifications 1457 2i8 6-3 Am. 1

B. TMI Plant Operational Review

• Topics which provide information covering the plant general operating characteristics and specific procedural guidance address the following areas:
• Heat Transfer and Fluid Dynamics Principles of Reactor Operation and Reactor Theory
• General and Specific Operating Characteristics
• Administrative Procedures, Conditions and Limitations
• Fuel Handling and Core Parameters Presentations on plant operation are designed to give detail-ed information on fundamental plant operation and specific procedural guidance. The specific topics are:

1. Heat Transfer and Fluid Dynamics

2. Reactor Theory

3. Use of Procedures

4. Operating Characteristics Review-including natural circulation

5. Solid Plant Operations

6. Operational Chemistry

7. Standard and Emergency Operating Procedures-(covered in nine sections)

(1) Administrative Procedures (2) Limitations and Precautions (3) Emergency Procedures (4) Emergency Feedwater Procedures (5) Reactor Coolant Pump Procedures (6) Electrical Power Emergency Procedures (7) Primary System Leak Emergency Procedures (8) Operating Procedures (9) Steam System Emergency Procedures

8. Technical Specifications - Limiting Conditions for Operations

9. Technical Specifications Review

10. Fuel Handling and Core Parameters

11. NRC Prompt Notification Enforcement Policy 1457 219 C. Radioactive Materials Control Topics which provide information covering radioactive ma-terials control address the following areas:

6-4 ^** I

• Radiation Control and Safety
• Radioactive Material Handling, Disposal and Hazards
• TMI Emerger. .y Plan The specific topics are:

1. TMI Radiation Emergency Plan

2. Radiation Safety and Radioactive Materials Control

3. Radiation Monitoring

4. Radioactive Waste Disposal

5. Liquid and Gaseous Releases D. TMI Plant Transient Analysis Topics which provide information covering plant abnormal operating characteristics and plant transients address the following areas:

• TMI-2 ' snsient Safety Analysis for TMI-l
• TMI Simulator Training The specific topics are:

1. TMl-2 Transient

2. Small Break Loss of Coolant Accident Operator Guidance

3. Reactor Coolant System Elevations and Manometer Ef fect

4. Expected Instruments and Plant Response to Transients

5. TMI Control Room Session

6. Safety Analysis Workshop In addition to these topics, specifically designed training sessions were conducted at the Babcock and Wilcox Simulator Training Center. These training sessions involved discussion of plant transient information and simulator training ses-sions where specific casualty situations were handled by the trainees.

The topics covered included:

1. Power Distribution and Rod Withdrawal Limits

2. heat Transfer and Fluid Flow

3. Small Break Analysis 4 Safety Analysis

5. Unannounced Casualties (conducted on the simulator)

6. Special program on the B&W Simulator covering the TM1-2 accident 6.4 PROGRAM RATIONALE The selection of topics to be included in the Operator Accelerat-ed Retraining Program was _ based on several factors. During the program formulation stage, the extensive training curriculum the TMI-l Reactor Operator and Senior Reactor Operator have already 1457 220 6-5 Am. 1 l

completed was balanced with the training needs related to the current TMI-1 and TMi-2 plant status. Specific sources utilized in identifying program topics include the following areas: A. Standard reierences for operator training programs considerea in determining course content include: 1.10 CFR 55 - Operator's License

2. NUREG-0094 - NRC Operator Licensing Guide

3. TMI-I FSAR

4. TMI-l Operator Requalification Program The topics included in the Operator Accelerated Retraining Program provide for coverage of all the areas in the NRC operators written examination (10 CFR 55.21/22). In addition topics included in the program include lecture requirements in the TMI Requalification Program (10 CFR 55 Appendix A and TMI-l FSAR Section 12).

B. Other Licensed Nuclear Operator Training References In making specific topic selections for the course content, other information sources for operator training were used. These sources include:

1. NRC Bulletin 79-05, 79-05A, 79-05B and 79-05C.

2. Metropolitan Edison Company commitments on operator training (J. Herbein letter to NRC dated June 28, 1979).

3. NRC letter - Order and Notice of Hearing, August 9, 1979.

4. Selected training programs conducted at other Babcock and Vilcox incident nuclear plants since the TMI-2 incident.

5. Interviews with TMI Operators.

6. TMI-l plant modifications (Short Term and Long Term).

7. TMI-2 incident information and other relevant License Event Reports.

8. NUREG - 0578 TMI-2 Lessons Learned 6.5 INSTRUCTIONAL PROCEDURE The Operator Accelerated hetraining Program topics are presented using a variety of instructional techniques. Instructional techniques utilized for particular program topics are selected to build comprehension of nuclear plant fundamentals, develop the 1457 221 6-6 Am. 1

ability to analyze and respond to plant operational situations , and ensure understanding of current TMI-l plant conditions and procedural guidance. In order to achieve the retraining program goals , the instruc-tional techniques utilized will include.

• Classroom Lectures
• Classroom Discussions
• Classroom Working Sessions
• TMI Control Room Training Sessions
• Nuclear Plant Simulator Practice Sessions (B&W Simulator Training Center)

A. Classroom Sessions In preparation for the classroom presentations conducted at IMI, an extensive program development process was completed. This preparation included the involvement of a primary and backup instructor for designated training sessions. Com pre-hensive lesson plans developed for the training sessions ensure a well directed approach for the presentations.

1. Topic Lesson Plan Preparation Lesson plans developed for the training sessions are in accordance with a standard format which includes all the elements of a comprehensive presentation and written guidance for carrying out a topic presentation.

Primary instructors assigned to prepare topic lesson plans have technical expertise in the specific areas covered by assigned topics. The primary instructor identified specific lesson plan objectives and developed the lesson plan material. Backup instructors assigned to assist in preparing topic lesson plans have experience in developing technical training material. In addition to assisting in topic lesson plan develop-ment the backup instructor also completes a Lesson Plan Dev elopment Summary which identifies essential informa-tion pertinent to the topic objective , instruction techniques , and etaluation procedures. The combined development ef forts of the primary and backup instructors is reviewed by designated training de pa r tment staff members at various stages to ensure a well directed , comprehensive topic presentation is adequately supported. 6-7 Am. 1 1457 222

2. Topic Classroom Presentation Classroom sessions are conducted following the direction provided by the topic lesson plan and lesson plan dev elopment summary. In order to ensure a comprehensive coverage of essential information in the classroom presentation, at least two people will be involved with the presentation. The primary instructor (or a desig-nated alternate) will present the topical information.

The backup instructor (or a designated alternate) will site in on the presentation and ensure that the essen-tial topic inf ormation is covered during the presenta-tion. This may involve clarifying certain points and a: king specific questions related to the topic lesson objectives and support material. In preparation for the classroom presentations, practice sessions involving the primary and backup instructor (or desige.ated alternates) are conducted as required. The practice sessions involve discussion of lesson material and presentation techniques and may include an abbreviat-ed practice presentation of part of the lesson. The practice sessions serve as a means of ensuring that actual lesson presentations will meet required standards and facilitate the achievement of the lesson objectives. The required extent of the practice session will depend upon the experience level of the primary instructor in presenting similar training material. B. Control Room and Simulator Sessions The Control Room and Simulator Training sessions are designed to enable hands on application of guidance provided to TMI-l operators. In preparation for these sessions, specific areas of coverage were designated to ensure essential items identi-fled and/or demonstrated for the operators.

1. Control Room Sessions A review with the information/ instrumentation available in the TMI-l Control Room is addressed in a specific session. This supplements the references made during other topic presentations which interfaced with Control Room features. A tour of the Control Room conducted under the guidance of a lesson plan prepared by a primary backup instructor team is designed to build the association of operational concept and guidance with actual system controls.

2. Simuistor Sessions The B&W Simulator Training is included in the program to provide actual practice for the TMI operators in handling plant transient si t ua tions .

                                                                   }}}

6-8 Am. 1 l 0

The training practices used during the simulator train-ing sessions enabled the following:

• Detailed use of procedures (including follow-up actions)
• Plant casualties carried out until a stable condition is reached
• Multiple plant casualties simulated
• Watch section members handling casualties ar, a team, with specific job assignments made
• Casualty conditions analyzed wish watchstander input, supervisor deciding course of actt'n :nd supervisor directing recoves
• Watch section members evaluated as a team on specific casualty response 6.6 EVALUATION PROCEDURE The Operator Accelerated Retra'. .irg Program is evaluated formally and informally in several mannets. Continuous informal evalua-tion is occurring during the training sessiosn as the instructor -

and/or backup instructor gauge trainee unoerstanding i asking questions and observing performance. Formal evaluations of the training program, instructor delivery, trainee performance and trainee knowledge level are also corduct-ed and analyzed. In addition, performance standards are speci-fied for key evaluation processes. A. Trainee Evaluation of the Program At the completion of each week of the training program, the trainees are asked to evaluate and comment on the training sessions. This evaluation encompasses the instructors, training materials, presentation techniques, and classroom facilities. Results of these evaluations are a means of measuring the trainees reaction to the training program. Problems which are identified by these evaluations are considered and resolved by the TM1 Training Department staff. Necessary changes to the program are factored into subsequent presentations. If a deficiency is deemed to be severe and cannot be otherwise compensated for, parts of the program will be repeated with the appropriate modifications incor-porated. B. Presentation Evaluations Each session of the program will be monitored and evaluatei by the session backup instructor or a designated alternate. 1457 224 6-9 Am. 1 l

An Instructor Evaluation Form is completed for the session and a presentation grade computed. To ensure the overall quality of instruction for each session, the following minimum standards are established.

1. Individual Presentation Standard Presentation Grade 2,2.5 (on a 4.0 scale)

The Presentation Grade is the average grade of all the individually graded entries on the Instructor Evaluation Form.

2. Topic Presentation Standard Topic Grade 2,3.0 (on a 4.0 scale)

The Topic Grade is the average grade of all the indi-vidual presentation grades for the topic. Presentations which do not meet the minimum standards will be subjected to the following:

1. Weaknesses found in the presentation will be discussed with the instructor.

2. Key concepts which are not adequately covered in this presentation will be presented again to the trainees in a subsequent training session.

3. Trainee performance on quiz questions on the concepts covered in the presentation will be evaluated. If trainee performance of 70% is found, the entire training session will be repeated for the af fected trainees.

C. Knowledge Evaluations by Quiz Each lesson plant for the program is developed with represen-tative quiz questions identified. During each week of training, quizzes will be administered and utilized for evaluation of trainee knowledge level. The quizzes will meet or exceed the following quiz standards:

1. Quizzes will be administered each week.

2. Each quiz will consist of at least ten questions.

3. At least 75% of the individual lesson plans presented during the week will have representative questions included in one or more of the quizzes.

4. A variety of question types may be used, but essay questions will predominate. Predetermined quiz question point values will be aasigned f 2: evaluation purposes.

1457 225 6-10 Am. 1 f

Quizzes will be scored and a grade for each quiz determined. To ensure satisfactory level of understanding of the weekly program material, the following minimum standard is es-tablished for each trainee's performance:

1. Individual Quiz 5tandard Individual Quiz Grade > 80%

For trainees who do not meet this standard, the following will occur:

1. Trainee will review the program the program material by reviewing the topic lesson plant and/or handouts.

2. Trainee will review the material with a designated staff member

a. Control Room Operators and Shif t Foreman will review the material with the Shif t Supervisor.

b. Shift Supervisors and licensed plant management will review the material with a designated instructor.

3. Another quiz will be administered and graded with the same standards in effect. The quiz will cover the material included in the unacceptable quiz (zes) and will be composed of questions not previously used during the program.

D. Knowledge Evaluation by Oral and Written Comprehensive Examination

1. Following the completion of the program, an Auditor Group will conduct a written and oral evaluation of the licensed trainees. The evaluation will be equivalent to an NRC administered licensing examination. It will include an expanded examination section covering the Operator Accelerated Retraining Program objectives.

Each successful trainee will be reuqired to pass the audit examination with the minimum examination standard.

2. Licensed Unit 1 personnel who have successfully complet-ed the Operator Accelerated Retraining Program will finally be required to take an NRC administered oral and written license examination.

6.7 PROG W FOR>1AT The Operator Accelerated Retraining Program is developed in over sixty individual lessons involving classroom presentations, TMI Control Room walkthrough and simulator training sessions. The entire program is scheduled for completion in seven modules, with a module cosisting of 4 to 5 days (8 hr/ day) of training. 6-11 Am. 1 l

Structuring the program into modules enable the scheduling of the presentations to occur during the six weeks cycle TMI training shift, or as a full time program. The content of each module is a selected grouping of individual lesson plans which cover material which is related to similar subjects. The modules are identified in Appendix B and are representative of the program scheduling. A. Simulator Training Module The initial program training module involved four and one-half days of training at the Babcock and Wilcox Nuclear Training Center. The module content included classroom training sessions and Control Room operational sessions. The indivi-dual topics were:

1. Power Distribution and Rod Withdrawal Limits (4 hours)

2. Heat Transfer and Fluid Flow (4 hours)

3. Small Break Analysis (4 hours)

4. Safety Analysis (4 hours)

5. TMI-2 Accident Analysis (4 hours)

6. Unannounced Casualties (16 hours)

The plant casualties included:

a. Natural Circulation Cooldown

b. Total Loss of Feedwater with no Emergency Feedwater (TMI-2 Accident)

c. Station Blackout (with diesels)

d. Los; of Coolant Accident

e. Steam Generator Overfeed

f. Steam Generator Tube Leak

g. Steam Leak in the Reactor Building The simulator training module provides an overview of guid-ance for operators which has resulted from analysis of the TMI-2 incident and involvement in simulated plant abnormal and emergency conditions. This initial program module supplemer ced previous operator training and provided a reference point for subsequent program modules dealing with detailed plant systems, operator guidance and nuclear plant fundamentals.

1457 227 6-12 Am. 1 1

B. TMI Module One The first module of the program conducted at TMI involved four ays of classroom training focused on nuclear plant fundamentals intergrated with specific plant operational characteristics. The individual topics are:

1. Heat Transfer and Fluid Dynamics (16 hours)

2. Reactor Theory (16 hours)

The content of module one provides an in-depth coverage of the fundamental aspects of nuclear reactor control and nuclear reactor heat remov al . These topics review principles necessary for understanding the purpose and function of nuclear plant systems, operational procedures and required operator actions for safety operating TMI-1. C. TMI Module Two The second module of the program conducted at TMI involves three and one-half days of classroom training covering specific IMI-l plant information on selected plant transients, plant systems and the Radiation Emergency Plan. The indivi-d al topics are:

1. TMI-2 Transient (4 hours)

2. Reactor Coolant System (5 hours)

3. Make-up and Purification System (4 hours)

4. In-Core Instrumentation (I hour)

5. Control Rod Drive System (4 hours)

6. Nuclear Instrumentation (2 hours)

7. Integrated Control System (4 hours)

8. Radiation Emergency Plan (4 hours)

9. NRC Prompt Reporting Requirements and Enforcement Policy (0.5 hours)

The content of module two provides detailed coverage of the TMI-2 Transient shich occurred March 28, 1979. This puts into perspective the plant systems and procedural training sessions included in subsequent program lessons. Detailed plant systems coverage begins in module two with sessions on key primary plant systems. 1457 228 6-13 Am. 1 l

D. TMI Module Three The third module of the program conducted at TM1 involves four and one-half days of classroom training covering spe-cific TMI-l plant systems and operational procedures. The individual topics are:

1. TMI-l Short-Term Modifications (4 hours)

2. Decay Heat Removal System (1 hour)

3. Decay Heat Closed Cooling System (1 hour) 4 Decay Heat River System (1 hour)

5. Core Flood System (1 hour)

6. Containment Isolation (1 hour)

7. High Pressure Injection (1 hour)

8. Use of Procedures (2 hours)

9. Nuclear Service Closed Cooling System (1 hour)

10. Nuclear Services River Water System (1 hour)

11. Reactor Building Emergency Cooling System (1 hour)

12. Intermediate Closed Cooling System (1 hour)

13. Feedwater System (1 hour)

14. Condensate System (1 hour)

15. Procedure Review-Reactor Coolant Pump Procedure (2 hours)

16. Emergency Feedwater System (2 hours)

17. Procedure Review-Emergency Feedwater Procedure (2 hours)

18. Main Steam System (1 hour)

19. Electrical Distribution (3 hours)

20. Emergency Diesel (2 hours)

21. Procedure Review-Electrical Power Emergency Procedure (2 hours)

22. Engineered Safeguards Actuation System (4 hours) 1457 229 6-14 Am. 1

The content of module three provides detailed coverage of selected TMI-1 primary and secondary plant systems. The systems covered in the program include systems essential to normal and emergency cooling of the reactor. E. TMI Module Four The fourth module of the program conducted at TM1 involves four and one-half days of classroom training covering spe-cific TMI-l plant systems, operational procedures and radio-active materials monitoring / control. The individual topics are:

1. Procedure Review-Primary System Leak Emergency Pro-cedure (2 hours)

2. Procedure Review-Steam System Emergency Procedure (2 hours)

3. Reactor Protection System (4 hours)

4. Operating Characteristics Review including Natural Circulation (4 hours)

5. Solid Plant Operations (2 hours)

6. Procedure Review-Emergency Procedure (2 hours)

7. Procedure Review-Operating Procedures (4 hours)

8. Radiation Safety and Radioactive Materials Control (4 hours)

9. Radiation Monitoring (4 hours)

10. Radioactive Waste Disposal (4 hours)

11. Liquid and Gaseous Releases (2 hours)

12. Operational Chemistry (2 hours)

The content of module four provides detailed coverage of selected TMI-l systems and plant procedures. Specific attention is given to normal and abnormal plant operations characteristics and related procedural guidance. Radiation safety, radiation monitoring, and radioactive materials control is covered to review existing guidance and present modifications made at TMI following the TMI-2 incident. F. TMI Module Five The fifth module of the program conducted at TMI inv olv e s five days of classroom training covering specific TMI-1 plant i457 230 6-15 Am. 1 l

systems, operational procedures, technical specifications and plant operational characteristics. The individual topics are:

1. Ventilation (3 hours)

2. Hydrogen Recombiner and Hydrogen Purge (1 '..our)

3. Technical Specifications-Limiting Conditions Opera-tion (4 hours)

4. Technical Specifications-Definitions and Safety Limits (2 hours)

5. Procedure Review-Administrative Procedures and Limita-tions and Precautions (2 hours)

6. Technical Specifications Review (4 hours)

7. Non-Nuclear Instrumentation and Interlocks (4 hours)

8. Small Break Loss of Coolant Accident Operator Guidance (4 hours)

9. Expected Instrument and Plan. Response to Transients (4 hours)

10. Reactor Coolant System Elevations and Manometer Ef fects (2 hours)

11. Fuel Handling and Core Parameters (4 hours)

12. Simulated Transients in Control Room (4 hours)

The content of module five provides detailed coverage of selected TMI-l Systems and plant procedures. Specific attention is given to normal and abnormal plant operating characteristics and related procedural guidance, including plant technical specifications. The TMI-l Control Room is used to develop further relationship between expected plant response to operational situations and actual control instru-mentation locations and features. G. TMI Module Six The sixth module of the program conducted at TMI involves five days of classroom training covering TMI-l plant modifi-cations and extensive coverage of safety analysis for TMI-1. The individual topics are:

1. Computer and Computer Modifications (4 hours)

2. TMI-l Long Range Design Modifications (4 hours)

3. Safety Analysis Workshop (32 hours) 1457 23l' 6-16 Am. 1

The content of module six provides an overview of specific changes being planned and accomplished at TMI and provides an in-depth presentation of key safe'y analysis areas and their implication to TMI-l plant operation. The safety analysis training will cover several areas of integrated TMI-l plant response to normal and abnormal events and provide guidance in evaluating plant performance in real time. The fundamen-tal principles of plant operation and plant system informa-tion will be combined with existing plant data to analyze several categories of potential abnormal operating conditions and categories of plant emergencies. 1457 232 6-17 Am. 1

n D O Z w 1457 233 . _. . A

7.0 RADWASTE MANAGEMENT 7.1 General The purpose of this section is to demonstrate that (1) the decon-tamination and restoration of TMI-2 will not affect the safe opera-tion of TMI-1, and (2) that TMI-1 has adequate waste treatment, radiation protection, and sampling capabilities for the safe operation of the Unit independent of TMI-2. The separation or isolation of TMI-1 and TMI-2 is addressed specifi-cally with regard to the following topics.

a. Radioactive waste transfer piping.

b. Fuel handling building environmental barrier.

c. Liquid wastes.

d. Miscellaneous waste evaporation.

e. Solid waste disposal.

f. Sanitary facility drains.

g. Radiation Protection and decontamination area.

h. Nuclear sampling station and radiochemical laboratory.

1. Industrial Waste Treatment Facilities.

The adequacy of radioactive waste treatment, radiation protection, nuclear sampling and effluent monitoring facilities is addressed in the following supplemental topics:

a. Radwaste capability.

b. Plant shiciding. -

c. Auxiliary building ventilation.

d. Nuclear sampling.

7.2 Separation and Isolation of the Units 7.2.1 Radioactive waste transfer piping At the time of the accident at Unit 2, there were five system inter-connections between Unit I and Unit 2 which permitted the movement of contaminated or potentially contaminated fluids from one unit to the other. In addition, a sixth intertie was planned. These interties are shown on Figure 7.3 Sheets 1 and 2 and are dis-cussed in the following subparagraphs. 7-1 Am. 4 1457 234

7.2.1.1 The contents of the Unit 2 Reactor Coolant Bleed Holdup Tanks may be transferred to the Unit 1 Reactor Coolant Waste Evaporator through the first interconnection shown on Figure 7.3. (heference GAI Dwg. C-302-692 and TMI-2 FSAR Figures 11.2-1 ano 11.2-2.) 1.2.1.2 Condensate from the Unit 1 Miscellaneous Waste Evaporator may be transferred to Unit 2 Evaporator Condensate Test Tanks via the second interconnection shown. (Reference gal Dwg. C-302-692 and TMI-2 FSAR Figures 11.2-1 ano 11.2-2.) 7.2.1.3 Us?.ng the third interconnection, the contents of the Unit 2 Neu-tralizer Tanks, Contaminated Drain Tanks, Reactor Coolant Bleed Holdup Tanks, Auxiliary Building Sump Tank and Miscellaneous Waste Holdup Tank may be transferred to Unit 1 for storage and processing in the Unit 1 Liquid Waste Disposal System. (Reference GAI Dwg. C-302-692) Similarly Unit 1 miscellaneous waste may be transferred to the Unit 2 Miscellaneous Waste Holdup Tank or Neutralizer Tank (Reference TMI-2 FSAR Drawing 11.2-2). 7.2.1.4 The fourth interconnection is shown on Figure 7.3 Sheet 2. This interconnection permits the movement of evaporator concentrates between units. Reclaimed boric acid as well as bottoms from the miscellaneous waste evaporator may use this path. (keference gal Dwg . C-302-6 92 and IMI-2 FSAR Drawing 11.5-1). 7.2.1.5 The final existing interconnection permits movement of spent ion exchange resin between units. (Reference GA1 Dwg. C-302-693 and TMI-2 FSAR Drawing 11.5-1.) 7.2.1.6 The connection permitting transfer from the Unit 1 keclaimed Boric Acid Tanks to the Unit 2 Concentrated Waste Tank or keclaimed Boric Acid Tank was not installed. 7.2.1.7 The methods to be used to isolate the aforementioned interconnec-tions will be by disconnecting control air, removing f uses from control circuits and MCC's, installation of spectacle, blind or pancake flanges, disablement of manual valves or reconnection to special facilities being installed for Unit 2. The specific method selected for each is designated on Figures 7.3 and 7.4 and will be reviewed, approved and incorporated into operating procedures. These lines will not be severed. 7.2.1.8 Subseq uent to the accident at Unit 2, the Auxiliary building Emer-gency Liquid Cleanup System (Epicor II) was installed in the Chemical Cleaning Building (CCB) for the purpose of processing the liquia radwaste accumulated in Unit 2. Processea water may be discharbec from the CCB by Transfer Pump ALC-P-5 to the locations shown on Figura 7.4 which is a simplified Flow Diagram of the Transfer Pump discharge circuit. As is shown on this figure, the processec water may be:

a. discharged to a truck at a station next to the CCs.

b. discharged to the Unit 1 Waste Evaporator Concensate Storage Tanks, or 7-2 Am. 4 1457 235

c. discharged to the Unit 2 Evaporator Condensate Test Tanks directly, or via the Evaporator Condensate Demineralizers.

Inadvertent discharge to Unit I components is virtually precludea by extensive isolation:

a. ALC-V169 is locked closed preventing discharges to Unit 1.

b. WDL-V421 and WDL-V422 a te locked closed to prevent discharges to individual components in Unit 1.

c. A pancake fitting is also installed at a spool piece in the transfer line to prevent discharges to Unit 1.

d. An existing line which was tapped to get the processeo ca.ter through the yard has a tee in Unit 1. A blind flange is on the unused branch of this tee.

7.2.2 Fuel Handling Building Environmental Barrier Due to the physical layout of the Units thare is a common air space connecting the individual units' Fuel Handling building. The common air space extends to the primary personnel access from the TMI-1 radiation protection control point to the TMI-1 Auxiliary buildiag. The original conceptual design for the separation of the units requirea the installation of a physical barrier wall that would extend to the ceiling of the fuel handling building. The wall would constitute a barrier to unrestricted personnel access and communication of the ventila-tion systems of the individual units. Upon close examination during the engineering of the barrier, it was determined that the concept was impractical because of the proximity to safety related equipment, (2) limitations to the use of the f uel building crane needed for fuel receipt and (3) periodic temporary breeches in the wall during transfer of the crane from unit to unit. An alternate concept that is currently being studied involves a physical isolation of the Unit 1 Auxiliary building ano f uel building access way at the 305' elevation from the remainder of the f uel handling building. The air space of the fuel hanaling buil ling operating levels would be common with ventilation modi-fications made to minimize the communication of air between the units. An approved environmental barrier system will be f unctional in the fuel handling building prior to start up of TM1-1. 7.2.3 Liquid Radwastes and Miscellaneous Waste Evaporator The original design of the TMI-1 liquid radwaste treatment system will in no way be impaired by the separation of the units. The TMI-1 radwaste system was cesigned to conservatively 7-3 Am. 4 1457 236

handle the waste generated from a sinbl e ope;ating unit and IMI-1 operated successf ully in the single unit mode in excess of three years during TMI-2 construction. The TM1-2 design utilizec the conservation of the TMI-1 system to provide radwaste treatment capabilities for liquid racwaste generaten from its operation. Therefore, all TMI-2 liquid radwaste had to be transferred to IMI-1 for treatment. Therefore isolation of the units will increase the TMI-1 liquid radwaste capability relative to that available during the pre-accident pe riod . The Miscellaneots Waste Evaporation is the primary prccess equipment used to treat liquic raawaste. The evaporator has been used successfully in the past for a single unit waste treatment requirements. Because of interferences in TMI-1 caused by the storage of TMI-2 waste waters, the evapora-tion is not currently in use. At the time of startup of TMI-1 (or within the month previous to starting) the evaporation will be shown to be operational or replacement equipment shall be available to supplement evaporation operation. Although, TMI-1 radwaste processing capabilities are currently limited by the presence of TMI-2 liquid radwaste stored in TMI-1 process vessels, this limitation will be completely removea by transferring and flushing all TMI-2 waste stored in TMI-1 to TMI-2 facilities prior to startup. TMI-2 miscellaneous liquid radwaste will be processed in systems installed (or to be installed) in accordance with the recovery effort. The systems include Epicor 2, the submeto:' 'emineralizer system and an evaporator / solidification system. A complete description of the TMI-1 radwaste treatment system capability is provided in section 7.3.1.1.1 of this report. 7.2.4 Solid Waste Disposal Prior to the accident, TMI-2 radwaste system was completely dependent on TMI-1 facilities for solid waste processing and disposal of evaporation concentrates, spent resin and compactible trash. Isolation of the two units will enhance the ability of TM1 to process solid radwaste. IMI-2 had installed separate trash compact-ing equipment and will obtain the equipment to solidify liquic anc wet solid waste as part of the recovery progran. TMI-1 capabilities are not required for the TMI-2 recovery program. TMI-1 currently has an opeating trash compactor used exclusively for TMI-1 trash. Solicification facilities will be available through the use of a contractor for the short term and subsequently through the use of a permanently installed system. Solidification capabilities are discussed more thoroughly in Section 7.3.1.3. 7.2.5 Sanitary Facility Drains By design, sewage and sanitary drains from TMI-2 would join those f rom TMI-1 at the sewage pumping station and then be treated in a common sewage treatment plant. Operationally, the sewage from the indivlaual unit currently remain separate 7-4 Am.4 1457 237

and is trucked offsite for disposal at a municipal sewage treatment facility. In the event that any TMI-2 sanitary facilities become contaminated with radioactivity, the ability to reaove TMI-01 sanitary waste from the site woula not be red uced . 7.2.6 Radiation Protection and Decontamination Areas TM1-1 has , installed as permanent plant equipment, all facilities necessary to support the raciation protection and cecontamination activities of a single unit. Any facilities in TMI-1, that in the past had been sharec by both units, will be used exclusively by TMI-1 subsequent to startup. Since the accident, TMI-2 has established a separate radiation protection organization utilizing dedie tted personnel and equipment. External facilities have been esta'e lished under the recovery program to replace facilities previously shared with TMI-1. 7.2.7 Nuclear Sampling and Radiochemistry Laboratory The Temporary Sample Sink System consists of tubing valves and other equipment necessary to satisfy all Unit ~I sampling requirements without utilizing the Unit I primary sample lab. Tubing and valves will be provided to facilitate sampling the following; Reactor Coolant Bleed Tanks, Miscellaneous Waste Holdup Tank, Mini Decay Heat System, Pressurizer Steam Space, Pressurizer Water Space, Reactor Coolant Letdown and the Fuel Pool Waste Storage System. The system will include all necessary tubing and equipment to allow adequate sample recirculation from the sample source back to that source. In 4.ddition, any purge required into the Sample Sink will be containaa in a drain cask at the sink. This cask will be pumped to the Reactor Coolant Bleed Tank when the level reaches a predetermined point. The system will also provide an incline Boron meter for measuring boron concentration in the samples drawn from the Unit 11 RCS. The Baronometer will be equipped with its own calibration equip-ment for periodic, on-line calibration. The entire sink will be shielded and enclosed with its own independent ventilation system which will exhaust into the Unit II Auxiliary Building Ventilation System. The TMI-2 Temp .ary Nuclear Sampling Facility is scheduled to be operational by January 1,1980. 7.2.8 Industrial Waste Treatment Facilities The Inductrial Waste Filter System (IWTS) and Industrial Waste Treatment System (IWTS) have been installed to camply with the 7-5 Am.4 1457 238

Environmental Protection Agency's NPDES program. These systems remove oils, greases, suspenced solids and trace quantities of copper and iron from secondary plant waste water. (Figure 7-1 indicates the flow paths for specific waste water sources.) The systems are common to ooth operating units and will continue to be used. Because of the increased possibility of radioactive centamination from TMI-2, the transfer of waste water to the facilities from TMI-2 will be performed under strict procedural control. Additions 1 procedural controls will be implemented when pumping from Unit I to the IWFS or ISTS shoula the TMI-1 secondary plant sumps become contaminatea. 7.3 Supplemental Topics 7.3.1 Radwaste Capability 7.3.1.1 Liquid Radwaste Processing 7.3.1.1.1 General Approximately 79,000 gallons of liquid radwaste related to the TMI-2 recovery is currently stored in TMI-1 process vessels. This liquia will be transferred to TMI-2 and the TMI-1 liquid radwaste processing system will be returned to a state of full operability prior to the TMI-1 startup. No system modifications are considered necessary to upgrade the system as a result of isolating the units. Because the tank used for storage of TMl-2 waste blocked the main process path to the Miscellaneous Waste Evaporator, that process unit has been out of service. The Epicor I system is currently in operation to treat TMI-1 wastes in lieu of evaporation and to process TMI-2 waste with low concentrations of radioactive material. Following the physical separation of the units low activity TMI-2 waste will be processed using TMI-2 facilities. Epicor I will be retained until both evaporators in TMI-1 have been demonsttut-d to be operable to within 75% of their nominal design capacities. Descriptions of the TMI-1 liquid radwaste processing system and the Epicor I system are provided in Section 7.3.1.1.2 and 7.3.1.1.3 res pec tively. 7.3.1.1.2 Liquid Waste System and Equipment TMI Unit 1 provides two strings of equipment for the processing of radioactive liquid wastes. One string of equipment is associatea with the processing of reactor grade water including spent fuel pool water and water on recycle through the decay heat system. The o.:her string of equipment is associated with the processing of miscellaneoas radioactive liquid wastes producea within the auxiliary and f uel handling buildings as a result of the processing of reactor and spent f uel pool liquids plus slightly radioactive liquids that occasionally occur in the secondary system as a result of steam generator tube leaks. The reactor grade water processing string incluces the following components: 7-6 am.4 1457 239

One 780 ft.3 (5,500 gallons) Reactor Coolant Drain Tank: for suppression and collection of pressurizer relief anc ccllection of process liquid f rom valve stems leak of f s in the reactor coolant system and from th~ reactor coolant system and from the reactor coolant pump seals. Two 11,000 ft.3 (80,000 gallons) Reactor Coolant Bleed Tanks: for the collection of letdown from the reactor coolant system and accept condensate from the seconaary system. L ther of these tanks can also be used for injection of boric acid solution into the reactor coolant system. One 11,000 ft.3 (80,000 gallons) Reactor Coolant Bleed Tank: for injection of feed solution into the reactor coolant system. This tank can also accept letdown from the reactor coolant system. Two 150 gpm precoat filters: for removing suspended anc ionic solids from reactor grade water. One may be used for treating from miscel-laneous wastes. Two 70 gpm demineralizers (normally cation resiva only): for removing ionic solids from reactor grade water. One 12.5 gpm reactor coolant evaporator: for concentrating the reactor grade water for reuse and producing a purified distillate. Two 920 ft. 3 (6,500 gallons) reclaimed boric acid tanks: for storage of concentrated reclaimed reactor coolant grade water. The miscellaneous waste string provides the following items of equipment: One 3124 ft.3 (20,000 gallons) miscellaneous waste storage tank: for accumulating waste liquids from various sumps, vents and drains within the auxiliary and fuel handling buildings. One 666 ft. 3 (4,900 gallons) neutralizer feed tank: for accumu-lating and storing solutions to be neutralized. One 194 ft.3 (1,000 gallons) neutralizer tank and mixer: for neutralizing solutions from the regeneration of the ceborating demineralizer resins and adjusting the pH or adding antifoam agent to miscellaneous a'.d laundry wastes prior to their evaporation. One 779 ft. 3 (5,500 gallons) neutralized waste storage tank: for acenmulating neutralizec wastes prior to their evaporation. One 1,120 ft. 3 (8,000 gallons) spent resin stvrage tank: for the accumulation and storage of radioactive spent resin generated throughout the unit. One 590 ft. 3 (4,000 gallons) used '11ter precoat storage tank: for the accumulation and storage of use d precoat material. 7-7 .Am. 4 143/ 44d

one 12.5 gpm miscellaneous w.ste evaporator: for concentrating miscellaneous wastes for pacnaging and shipment off-site. Two 920 ft.3 (6,500 gallons) concentrated waste storage tanks: for the accumulation and storage of bottoms from the miscellaneous waste evaporation. The equipment provided that is common to both the reactor coolant and miscellaneous waste strings is: Two 15 pgm evaporator condensate demineralizers (mixed bed resin): for removing any borate and other ions that may have carriec over in the distillate of the reactor coolant or miscellaneous waste eva po ra to rs . Two 1,234 ft. 3 (8,400 gallons) evaporator condensate storage tanks: for accumulating, storing, and sampling the evaporator distillate prior to its recycle to the unit for reuse or discharge to the mechanical draft cooling tower effluent for disposal to the environment. Two waste transfer / disposal pumps: a flow monitor; radiation monitor with alarms and action set points on both the flow and radiation monitors. The only process equipment shared by the reactor coolant anc miscel-laneous waste prcccccing strings is one precoat filt.er , that may be utilized to filter miscellaneous waste in the event that this is required, and the ability to concentrate either waste stream in either evaporator in the event that one evaporator is unavoicably out of service for an extended period, both processing strings provide duplication of tanks , pumps and process equipment in all areas of high service factor in order to allow operating to proceed normcIly in the event that an item of equipment is unvoidably out of service for an extended period. Both processing strings provide numerous cross connects between storage tanks and alternate process paths to provide emergency or additional storage capability for various waste streams and versa-tility of treatment of the various waste streams, respectively. both process strings also provide the capability to recycle either the concentrated or the purified prod uc t produced from the waste in the event that it does not meet the quality standards set fcr it. The storage tank capacities and process flow rates were conserva-tively chosen with respect to the design basis annual liquid waste generation rate which assumes approximately three times the normally anticipated generation of reactor coolant grace water anc twice the normally anticipated generation of miscellaneous liquid wastes (refer to Table 7.3-1, f or a breakdown of the design basis quantities of waste liquids generated annually within the unit). 7-6 Am. 4 1457 24i

Further contributing to the availability anc security of the liquic waste system is the fact that all of the above equipment is located within Seismic Class I structures that hr;c oeen hardened to with-stand an aircraft impac t . Within these structures, all equipment that is anticipated to become a significant radiation source is housed within 2 to 3 foot thick shielo walls for the protection of plant personnel from radiation. The atmosphere of each of these shielded cubicles is maintained at a slightly lower pressure than that in surrounding areas to ensure that any racioactive gas leakabe is away from plant personnel. Based on the above indicated systems ano equipment, the design basis waste liquid quantities generateo annually and the design basis activity levels analysis indicates that the liquid effluent leaving the plant site is well within the limits specified by Appencix 1 with the exception of tritium. With 17,500 gpm of the cooling tower effluent allocated to Unit 1, the Unit 1 annual discharge volume for which dilution credit may be taken is 3.48 x 10 10 liters. The annual quantities of mixed fission products (excluding tritium) and tritium discharbed from Unit 1 are 49,000 uCi ana 5.02 x 108 uCi respectively. These design basis numbers result in an annual average mixed fission product concentration in the plant effluent of 1.4 x 10-6 uCi/ liter (compared with the Appendix I limit of 2X10-5 uCi/ liter) which ia about 1/14 the Appencix I limit; whereas the annual average concentration of tritium in the plant effluent of 1.45 x 10-2 uCi/ liter (compared with the Appendix I limit of 5 x 10-3 uCi/ liter) is about 3 times the Appendix I limit. 7.3.1.2 Waste Gas System 7.3.1.2.1 General The TMI-1 gaseous rad waste system is totally independent of the analogous system in TMI-2. The system has functioned satisfactorily to support the unit's requirements since initial startup. The Thl-1 waste gas system has had a leak tight ' .'. story since modifications were made during the 1976 ref ueling outage. The system's leak rate and component malf unctions frequency have been very low for the period subsequent to the 1976 ref ueling outage. In June 1979, a pressure drop test was conducted on the TMI-1 Waste Gas System. No indication of leakage was noted during the test. 7.3.1.2.2 Gas Waste Systems and Equipment The TM1-1 gas waste systems are:

1. The High Level Waste Gas System: for the accumulation, storage and re-use or controlled disposal of high activity level gases evolvec from primary coolant in various systems within the unit.

2. The Auxiliary anc Fuel Handling Huilding Ventilation Systems:

for the continuous particulate and charcoal filtration, moni-toring and disposal of small quantities of racioactive gases released to the atmospheres of the auxiliary and f uel hancling b uilding s . 7-9 Am. 4 1457 242

3. The Reactor Building Purge System: for continuous particulate and charcoal filtration, monitoring and disposal of small quantities of radioactive gases appearing in the reactor building atmosphere when it is perit dically purgec.

In combination, the above three systems retain the bulk of the radioactive gases generated within the station for a period of decay before release and continuously particulate and charcoal filter and monitor the ventilation releases from the plant to ensure that their activity burden remains within acceptable levels. A. High Level Gas Waste System The TMI-1 high level gas waste system consists of low pressure vent header that includes the gas spaces of seven tanks storing reactor coolant anc a 400 ft. 3 waste gas delay tank. Two 40 SCFM capacity waste gas compressors take suction from the outlet of the waste gas delay tank and compress the gas into one of three 1,125 ft 3 waste gas cecay tanks for a minimum of 30 days holdup for radioactive decay prior to disposal of the gases. The estimated decay period may be extencec to 90 days if required. At the 80 psig storage pressure, the three waste gas decay tanks can accommocate the storage of the equivalent of over 15,000 f t. 3 of gas at atmospheric pressure. The system provides for recycle of gases stored in the waste gas decay tanks. Except for re-use or disposal of radioactive gases, the vaste gas system operation is automatic. As the pressure within the low pressure vent header system rises above the high set point for it, the waste gas compressor compresses the excess into the waste gas decay tank until the low pressure vent heacer pressure falls below the high set point. In the event that the low pressure vent header falls below the low pressure set point , fresh nitrogen gas or gas from a filled waste gas decay tank is added to it by an automatic valve until the system pressure again exceeds the low pressure set point . Since there is about a 0.6 psi differential between the ldgh and low pressure set points of the low pressure vent heacer system, the system pressure may float within this band with no gas being either acded to it or removed from it. Re-use or disposal of gases in a waste gas decay tank is accomplished by operator action. In the event that the stored waste gas is considered acceptable for re-use (basically less than 3% hydrogen by voltne), it is valved upon to recycle its gas to the low pressure vent heacer system as called for by the automatic control valve in 7-10 Am. 5 1457 243

the gas addition system. In the event that the storea waste gas is to be disposed to the environment, it is sampled and analyzed. Basea on Unit Tech Specs. , the gas is then released fram the waste gas decay tank at a controlled rate between 1 to 10 SCFM, via raciation anc flow rate monitors, into the auxiliary building ventilation system upstream of the particulate and charcoal filters. The flow rate and t.ctivity level of the gas release are monitored both be' ore they enter the auxiliary building ventilation system and as they exit the controlled area via the units' vent stack. In the event that either the radiation monitor or the initial gas release flow rate device sense an "above set point" activity level or flow rate, respectively, the gas release from the cecay tank is automatically terminated. The low pressure portion (upsteam of the waste gas compres-sors) of the vent header system is protectea from over pressure by relief valves on the vent header itself anc by water filled Icop seals on the overflow of the Miscellaneous Waste Storage Tank. Tne loop seal also protects the tank from excessive vacuum that may dcvelop as the Miscel-laneous Waste Storage Tank or other system tanks are drained. The Waste Gas Decay Tanks are protectea from overpressures by individual relief valves as are the gas compressors. Listea below are the radwaste system compo-nents that are connected to the waste gas system. All other tanks in the liquid radwaste system (except the R. C. Drains Tank) are vented to the TMI-1 auxiliary builcing. COMPONENT RELIEF SET POINT DISCHARGE PolNT la . Vent Heaaer 8 psig Ventilation System before the filters Ib. Vent Header 8.5 psig (loop seal) Auxiliary Building (3.25 psig vacuum) Note: Reactor Coolant Bleed Tanks and the Miscellaneous kaste Storage Tank are directly connected to the Waste Gas vent he ader .

2. Gas Decay Tanks 85 psig Ventilation System before the filters

3. Misc. Waste 6 9 psig Vent Header RC Evaporation

4. Reclaimed horic (Existing Vent Vent deacer Acid Storage Header Pressure)

Tanks 7-11 Am. 4 1457 244

B. The Auxiliary and Fuel Handling Builcing Ventilation Systems and the Reactor building Purge System The auxiliary and fuel handling building ventilation systems and the reactor building purge systems serve primarily to distribute to and exhaust ventilation air from the areas indicated. 7.3.2 Plant Shielding 7.3.2.1 General: The purpose of this section is to identify the necessary action to be taken for providing additional shielding, procedures or plant modifications in vital areas and to equipment in which personnel occupancy may be unduly degraded by the large raciation fields in operating these systems in a post-accident environment. Systems, components, and areas considered subject to post-accident large radiation fields include: decay heat reraval, reactor building spray recirculation, make-up and let lown, waste gas, kad Waste Control Center, and Liquid and Gas Waste Panel. 7.3.2.2 Design Keview: In order to determine which areas and systems require additional shield-in or plant modifications, a review of existing design will be completed which will identify and recommend the cor-rective actions nceded in vital areas throughout the Unit. The review will encompass the accessibility and operability of the above stated systems and area in a post-accident environment. Bases to be used in comparing present design to post-acciaent environment. Bases to be used in comparing present design to po s t-a ccident radiation levels will be the fission product release as described in Regulatory Guides 1. 3 and 1. 4. The design review will be completed by January 1,1980. 7.3.2.3 Near Term Modifications: Those near term modifications, identified by the design review, will be completed prior to startup of Unit 1. 7.3.2.4 Long Term Modifications: Over the longer term modifications not comple ted before startup of Unit I will be completed by January 1, 1981. 7.3.3 Auxiliary Building Ventilation System 7.3.3.1 General The Unit-1 Auxiliary Building Ventilation System is combinea with the Fuel Handling Building Ventilation System prior to filtration 7-12 Am. 4 1457 245

and exhaust from the Unit. The Auxiliary building System is designed to maintain a 80,000 CFM average air flow with the Fuel Handling Building System adding on additional 40,000 CFM for a combined exhaust flow of 120,000 CFM. Each system has a separate set of supply f ans , AH-E Auxiliary Builcing and AH-E Fuel Handling Building, and the individual exhausts are monitored for racioactive particulate, iodine and gaseous activity, RM-A6-Auxiliary Building and RM - A4 - Fuel Handling Builoing, prior to combining into a common exhaust. The common exhaust flow is filtered by eight (8) parallel filter banks each consistinh of a Pre-filter, High Ef ficiency Particulate Absorber (REPA) anc Charcoal Adsorber. The exhaust is drawn through two of four exhaust fans , AH - E - 14A through 140, each sized to 50 percent of system capacity. The final fan exhaust is combinea and discharged thro ugh a sic le stack which is not common with any other building exhaust system. The exhaust is monitored for radioactive particulate, iodine and gaseous activity, RM - A8. There is no bypass capability around the filter units, resulting in all air during both normal and accident conditions being filtered. 7.3.3.2 Testing Requirements In that the Filter System, AH - F - 2A thru 2H, provide filtration for the Fuel Handling Building exhaust, the specifications for testing the Fuel Handling will apply to Auxiliary Building exhaust filters. The following specifications will apply:

a. At least,once per 18 months or after any structural maintenance on the HEPA filter or charcoal adsorber housings.

1. Verifying that the clean-up system satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a , C.5.c, anc C.S.d of Regulatory Guide 1.52, Revision 1, July 1976. The prerequisites of section 10.3 and 12.3 of ANSI-N510-1975 will not apply.

2. Verifying within 31 cays after removal that a laboratory analysis of a representative carbon sample obtainec in accordance with Regulatory position C.6.b of Regulatory Guide 1.52, Revision 1, of July 1976, meets the testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

b. Af ter every 720 hours of charcoal adsorber operation by verifying within 31 days af ter removal that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

c. Af ter each complete or partial replacemerc of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the D0D when they are tested in place in accordance with ANSI N510-1975.

7-13 Am. 4 1457 246

d. Af ter each complete or partial replacement of charcoal adsorber bank by verifying that the charcoal adsorbers remove > 99% of a halogenatea hydrocarbon refrigerant test gas when they are testec in-place in accorcance with ANSI N510-1975.

7.3.3.3 Implementation Schedule The testing schedule as previously describec will be in place prior to the restart of Unit-1. 7.3.4 Nuclear Sampling 7.3.5 Nuclear Sampling Capabilities 7.3.5.1 Post-Accident Sampling 7.3.5.1.1 General Purpose of this section is to identify the necessary action to be taken to facilitate the capability to perform timely sampling of Reactor Coolant and Containment Atmosphere in a post-accicent environment. Action to provide the capability to perform timely (within I hour) sampling of reactor coolant and containment atmosphere include: . a design review of existing systems anc component location, near term and long term modifications. 7.3.5.1.2 Design Review: To determine what moitfications are required, a design review of the existing systems and components will be completed. This design review will identify and recommend the corrective actions necessary to provide the capability to sample in a post-accident environment. The design review will encompass the accessability and operability of the reactor coolant sampling and containment atmosphere sampling sampling systems in a post-accident environment. The design bases to be used for this review will be the fission product release in an accident as assumed in Regulatory Guides 1.3 ano 1.4. The design review will be completea by January 1,1900. 7.3.5.1.3 Near Term Modifications: Those near term modifications, iaentified by the design review, will be completed prior to startup of Unit I. 7.3.5.1.4 Long Term Modifications: Over the longer term mo/ifications not completec before startup of Unit ' will be complete 3 by January 1,1981. 7-14 Am. 4 1457 247

7.3.5.2 Sample Drains The TMI-1 Nuclear Sampling sample and analysis drains are routed to the TMI-1 Auxiliary Building Sump to be processed by the Liquia Radwaste System. In the event of a plant accicent accident water including reactor coolant would be discharged to the sump as the result of sampling and analysis. Since the sump is not sealed radioactivity from the accident would escape uncontrollably into the Auxiliary building from the sump. To prevent this from happening, modifications will be made to pipe the radiochemical laboratory drains to the Miscellaneous Waste Storage Tank either directly or by way of an intermediate collection tank and pump (s). The Miscellaneous Waste Storage Tank would contain the laboratory wastes because the tanks gas space is connected to the vent header. Interneciate tanks and pumps that may be used would also be vented to the waste gas system. The laboratory waste collection modification will be operational by October 1, 1980. 7-15 Am. 4 1457 248

TABLE 7. 3.1 DESIGN RADIOACTIVE WASTE QUANTITIES Assumptions and Waste Source Quantity per Year, ft. 3 Comments I. Liquid Waste

1. Reactor Coolant System:

For selecting tank & equipment capacities 60,600 2 cold startups at beginning of core life; 3 cold startups thereafter at 77.5 days inter-vals, continuous shiu bleed during operation; boron re-moval via deborating demin-eralizers during last 55 days refueling. Total cycle time of 340 days. Therefore, full power operation for 335 days per year.

2. Sampling and Laboratory Drains 400 12 samples per week at 5 gal per sample.

3. Purification Deminer- 160 80 ft3 resin /yr. sluiced at alizers 2ft3 sluice water /ft3 resin.

4. Cation Deminerali-zers 288 144 ft3 resin /yr. sluiced at 2ft3 sluice water /ft3 resin.

5. Deborating Deminerali-zers 1,080 2 resin regenerations/yr. at 13.5 ft3 re

                                                  & rinses /ftgeneratingsolution resin.

6. Condensate Deminerali-zers 96 48 ft3 resin /yr. sluiced at 2ft3 sluice water /ft3 resin.

7. Precoat Filters 240 8 filter bed changes / yrs, at 30 ft3 sluice water / filter bed change.

1457 249

TABLE 7.3.1 (cont.) DESIGN RADIOACTIVE WASTE QUANTITIES Assumptions and Waste Source Quantity per Year, ft. 3 Comments

8. Misc. Sys. Leakage ,

Decontamination, etc. 70,000 1 gpm accumulation rate

9. Laundry (1) 7,300 150 gpd accumulation rate

10. Showers (1) 14,600 10 showers per day at 30 gal.

per shower. II. Gaseous Waste

1. Of f gas from reactor Degas at 25 cc H2 per liter coolant letdow 1,500 concentration.

2. Off gas from reactor Degas at 25 cc H2 per liter coolant sampling 10 concentration.

3. Makeup Tank gas inventory 1,000 Vent once per year.

4. Off gas from pressurizer 60 Vc.at once per year.

III. Solid Waste

1. Purification Resin 80 Resin replacement twice per yr.

2. Cation Demineralizers 144 Resin replacement four times pe r yr .

3. Deborating resin 4 One quarter of one resin bed replaced per yr.

4. Condensate Demineralizers Resin replacement four times resin 48 per yr.

5. Filter precoat 16 Filter bed replacement eight times per year at 2 ft3/repl.

6. Evapo "ator bottoms 811 (2)

Notes: (1) Only 10 percent of this is assumed to require processing in the liquid waste system. Remainder is discharged to sanitary waste. (2) Based on the following assumptions: 1457 250

TABLE 7.3.1 (cont.) DESIGN RADIOACTIVE WASTE QUANTITIES

a. Concentrate from 90 percent af item I.1 is reclaimed for rease. Remainder is concentrated by a factor of 20 for packaging (303 f t3).

b. Item I.2 is concentrated by a factor of 10 for packaging (40 ft3),

c. Items I.3, I.4, I.6, I.7 and 1.8 are concentrated by a factor or 500 for packaging (141 ft3),

d. Item I.5 is cencentrated by a factor of 10 for packaging (108 ft3),

e. 10 percent of items I.9 and 1.10 are concentrated by a factor of 10 for packaging (219 f t3).

1457 251

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8.0 SAFETY ANALYSIS

8.1 INTRODUCTION

Design changes af fecting the acceptance criteria for the TMI-l FSAR safety analyses arise from several sources. First is the TMI-l " Order and Notice of Hearing" (Reference 19) which contains NRC staff recommendations that certain changes be made to the plant. This order encompasses recommendations made in NRC bulletins 79-05 A, B and C and the TMI-2 Lessons Learned Task Force NUREG-0578 (Reference 20). Most of the changes listed below are being made in response to this order. Prior to the TMI-2 accident, B&W 177 FA plants received orders requiring modifications to the high pressure injection system to accommo-date certain small break LOCA's. These changes are being evalu-ated as well. A third source of changes has originated from plant upgrades that Metropolitan Edison believes would improve plant performance. Some of these modifications were being evaluated prior to the TMI-2 accident on March 28, 1979. Certain analyses have been, or will be, performed using the TM1-1 RETRAN computer model to evaluate the ef fect of the modified design on relev ant plant transients. These analyses will be selected in light of insight gained from the TMI-2 accident. The analyses of interest are:

1. The transition to natural circulation following a loss of of fsite power.

2. The feedline break accident with regard to functional require-ments for the emergency feedwater system.

3. Partial and complete loss of feedwater events and their sensitivity to: PORV setpoint, emergency feedwater flow, and reactor trip on loss of feedwater/ turbine trip. A second area of investigation was the PORV setpoint. As indicated in Reference 2, B&W analyses indicated that a setpoint of 2450 psig would prevent lif ting of this valve for all transients that have been experienced at B&W plants. The PORV setpoint of 2450 psig has been found acceptable based on the setpoint accuracy of the PORV and safety valves, and the frequency of calibration on the PORV actuation circuitry. Finally, RETRAN analyses will also be used to determine if overcooling can occur as a result of EFW operation following a loss of offsite power. The present design would allow 200% capacity if all three pumps operate and the OTSG 1evel would be controlled at 50% on the operate range.

These analyses will also support design decisions affecting plant modifications. A final source of input will come from the Abnormal Transients Operating Guidelines ( ATOG) Subcom-mi; tee of the B&W 177 FA Owner's Group. Met-Ed is a full participant in this group and will utilize results in developing operator guidelines , for TMI-1. The RETRAN model could be used in lieu of B&W analyses if plant specific analyses are required in developing operator guidelines. 1457 258 8-1 Am. 5

The decision to investigate additional accidents / transients was arrived at by reviewing the events in Table 8.2-1. These events were developed f rom the TMI-l FSAR, the Standard Format and Content Guide , Rev. 3, and NRC requests for additional informa-tion from the TMI-l and TMI-2 dockets. 8.2 AREAS OF INVESTIGATION The plant modifications which are being investigated are sum-marized below. They are grouped according to their origin. 8.2.1 Modifications Resulting from the August 9, 1979 Order

1. The reactor protection high pressure trip setpoint has been changed to 2300 reig from 2390 psig. This lower trip set-point in conjunc'. ion with the higher power operated relief valve (PORV) setpoint of 2450 psig results in a lower like-lihood of PORV operation.

2. A complete loss of feedwater flow will initiate a reactor trip.

3. A turbine trip will initiate a reactor trip.

4. The emergency feedwater system will be modified to allow

a. automatic initiation of the steam and motor drive EFW upon loss of all 4 reactor coolant pumps, feedwater/ steam differential pressure and a loss of main feedwater pumps. l

b. loading of EFW pumps on the diesel generators and dele-tion of the blackout start inte rlock.

c. alternate manual control for the EFW system.

5. The emergency feedwater system will be modified to start automatically on the following safety grade signals:

a. low steam generator level. Since this item is long term, plant safety will be discussed with and without these changes,

b. negative feed to steam dif ferential pressure.

c. loss of all four reactor coolant pumps.

8.2.2 Modification as Result of Order of May, 1978 Modifications to the high pressure injection system. The HPI injection lines have been cross connected to assure acceptable results from a break in a high pressure injection line. Cav ita t-ing venturis have been added to provide the proper flow split in the event of an HPI line break.

                                                                         }g 8-2                                Am. 5

8.2.3 Modification Originating from within Met-Ed

1. Post accident instrument and valve operator availability will be improved by the addition of heat shrink tubing.

2. Th- switchover of the ECCS system suction supply from the borated water storage tank (BWST) will be accomplished au".omatically rather than by operator action.

3. The reactor building spray system will be modified to delete sodium thiosulfate. Sodium hydroxide will be retcined.

This change will provide equal drawdown of the EkST and NaOH tanks for a large spectrum of single failures.

4. The fuel handling building, which is presently shared between TMI-1 and 2, will have an airtight barrier partitioning the building.

8.M.4 I&E Bulletin 79-05C Met-Ed is in the process of evaluating the response to this bulletin. It is expected that a reactor coolant pump trip will be initiated on a SFAS coincident with an indication of a large (in excess of 10-20%) void fraction. This or any other change will be evaluated with regard to their ef fect on the plant accident and transient analyses and plant operating guidelines. 8.3 EFFECT OF CRANGES ON SAFETY ANALYSIS Following are summaries of the accidents listed in Table 8.2-1. Table 8.2-1 indicates whare FSAR analyses took credit for non-safety grade equipment, or where mitigation is dependent on a specific operating / emergency procedure or design margin. These conclusions will continue to be revised to account for plant design changes. The event description and mitigating equipment are for the plant design before modification. The modifications discussed in the previous secitons were considered in the review of each accident. If a modification affected that analysis, then a note as to its safety significance was made under the " conclusions" section. 8.3.1 Rod Withdrawal from Startup (FSAR Section 14.1.2.2)

1. Description Uncontrolled reactivity excurcion starting from a suberitical condition of 1%ak/k at hot standby.

2. Acceptance Criteria

1. Limit power to design overpower (112%)

1457 260 8-3 Am. 3

11. RCS pressure not to excecd code allowable of 2750 psig.

3. Mitigation

1. RPS trip on high pressure for fast power rises, ii. Pressurizer code safety salves lif t and peak pressure is limited to 2515 psia.

iii. Doppler coef ficient provides a negative reactivity addition.

4. Conclusion The FSAR analysis still bounds the modified TMI-1 plant design. The RCS high pressure trip is lower and safety margins are increased. Since no credit was taken for opera-tion of the PORV, raising the valve setpoint does not change the analysis results. As discussed in Ref. 2, the PORV would lift for the worst case rod withdrawal accident which was analyzed in the FSAR. Nevertheless, the probability of occurrence has been decreased so that safety margins have been improved and lif ting of the PORV is not likely for a broad spectrum of rod withdrawal accidents.

8.3.2 Rod Withdrawal at Power (FSAR Section l's.l.2.3)

1. Description Accidental witudrawal of a control rod group at normal rated power , without ICS control and a 1% shutdown margin.

2. Acceptance Criteria

1. Limit power to design overpower of 112%.

ii. RCS pressure not to exceed code allowable (2750 psig).

3. Mitigation

1. RPS trips on high pressure for slow treaciants and high neutron flux for fast transients.

ii. Doppler and moderator coefficients provide negative reactivity addition.

4. Conclusions The FSAR analysis bounds the modified TMI-l plant design.

Lowering of the reactor trip setpoint increases safety margins for this event. Credit was not taken for PORV o pe ra tion. As discussed in Reference 2, some low worth rod 1457 261 6-4 Am. 3

withdrawals can result in PORV actua tion. Nevertheless, the probability of such an occurrence has been greatly decreased by the changes in the PORV and high pressure trip setpoints. 8.3.3 Moderator Dilution Accident (FSAR Section 14.1.2.4)

1. Description Diluted makeup water is inadvertently added to the reactor coolant system at a rate of 500 gpm beginning at normal powe r. RCS boron concentration is at its highest initial v alue . The result is a reactivity insertion, increased power, pressure and temperature. The addition of one makeup tank volume of unborated water changes the shutdown margin by

           .8% Ak/k-

2. Acceptance Criteria

1. Reactor power will be limited to less than the design overpower (112%).

11. Reactor coolant system pressure will be limited to less than code allowable 2750 psig.

iii. The minimum shutdown margin will be at least 1% ak/ k*

3. Mitigation

i. High pressure or high temperature trip.

ii. Termination of deborated water to makeup tank on reactor trip. iii. Termination of makeup flow on high pressurizer level.

4. Conclusion The FSAR analysis bounds the modified TM1-1 plant design.

Lowering of the high pressure trip setpoint incraases the safety margins for this accident. Operation of the PORV was not assumed in the original annlysis, and peak pressure is 2435 psia. Therefore, the PORV setpoint will not be reached during this transient. Reactor power is limited to 107.3%, and the final shutdown margin is greater than 1% a k/k even with the most reactiv e rod stuck out of the core all of the acceptance criteria for this accident are met. 6.3.4 Cold Water Addition (FSAR Section '14.1.2.5)

1. Description Startup of one or more idle reactor coolant pumps can cuase excess heat removal from the primary coolant system. This cooldown can cause positive reactivity insertions, w i h 8-5 A=. 3

res ult in a power rise. The worst case event is the startup of two reactor coolant pumps from 50% power. A tripped rod worth of 1% 4 k/k is used in the analysis.

2. Acceptance Criteria

1. Limit overpower to less than the maximum design overpower (112%).

3. Mitigation

1. RPS trip on high pressure for slow power increases or power / flow mismatch for rapid power increases.

ii. RC pump / power monitor limits initial conditions under which event can occur.

4. Conclusion Lowering of reactor trip setpoint increases safety margins for this event. The FSAR analysis was performed without taking credit for PORV. Peak pressure did not exceed 2400 psia, hence *.he PORV will not lif t during this event.

The FSAR analysis bounds the modified TMI-1 plant design. 8.3.5 Loss of Coolant Flow (FSAR Section 14.1.2.6)

1. Description Fuel rods experience a limiting DNB transient when all four reactor coolant pumps trip on loss of of fsite power or when one pump experiences a locked rotor resulting in an instan-taneous loss of flow. The loss of flow analysis is performed from 114% normal power, nominal reactor coolant pump flow, a

         +2 F core inlet temperature error and a -65 psi error in pressure. Reactor trip delay is assumed to be 620 ms. and a 1% Ak/k suberitical margin is assumed at hot standby. The ev ent is analyzed past the time that the minimum DNBR occurs.

The locked rotor accident is performed from an initial power level of 102% power, with a rampdown in flow from 100% to 75% in 100 ms. Temperature and pressure were the same as for the loss of flow accident. Reactor trip delay is assumed to be 650 ms.

2. Acceptance Criteria

1. DNB is greater than 1.3 for a loss of coolant flow.

ii. DNBR is greater tbta 1.0 for a locked rotor accident. 1457 263 8-6 Am. 3

3. Mitigation

1. Protection from four pump coastdown is by limitation of peaking factors, limitations on power level and the pump power monitor.

ii. Protection for the locked rotor accident is by the flux / flow monieor initiating reactor trip.

4. Conclusions The FSAR analysis for the four pump coastdown terminates prior to establishing stable decay heat removal by natural circulation. The EFW system will automatically start and maintain steam generator level at 50% on the operate range.

This design should result in the transition to stable condi-tions; a test will be perfromed to demonstrate this transition l prior to startup. 8.3.6 Dropped Control Rod (FSAR Section 14.1.2.7)

1. Description A dropped control rod reduces the average coolant tempe rature and reduces power. A return to full power may result in high local power density and heat fluxes. The analysis is per-formed at rated power with the most adverse values of the moderator and doppler coefficients (EOL) Rod worth are the maximum expected for full power operation with and without Xenon. Tripped rod worth is assumed to be 1% k/ k-

2. Acceptance Critpria

1. DNBR remains above 1.3.

ii. Reactor coolant system pressure is less than code allow-able (2750 psig).

3. Mitigation

1. The integrated control system inhibits withdrawal of control rods and ramps secondary side steam demand to 60%

rated power to prevent overcooling.

4. Conclusions This analysis has not been changed as a result of any of any TMI-1 plant design changes. Analysis results still show that the acceptance .riteris are met. 'c should be noted that while ICS action is assumed in this analysis, acceptable results are nuc dependent on ICS operation. The dropped control rod analysis performed in the TMI-2 FSAR does not assume ICS action, and demonstrates that the accident accept-ance criteria are met.

1457 264 8-7 Am. 5

8.3.7 Loss of Elegiric Power (FSAR Section 14.1.2.8)

1. Description Separation of the unit from the transmission network can result in the trip of the turbine and reactor. A core severe transient occurs if the ICS does not run tack the reactor load demand. The result is reactor trip on high pressure.

Cooldown is accomplished through the atmospheric dump or steam relief valves. la the presence of failed fuel and primary to secondary leaks, this event can lead to low levels of radioactivity release.

2. Acceptance Criteria

1. DNBR shall not be less that. 1.3.

11. Reactor coolant system pressure will not exceed coja allowable limits of 2750 psig.

3. Mitigation

1. Reactor trip on high pressure.

4. Conclusion This transient has an increased safety margin over the analysis performed in the FSAR as a result of the high pressure trip setpoint reduction to 2300 psig and the antici-patory reactor trip with turbine trip. In addition, a PORV setpoint of 2450 assures that the PORV will not be activiated (Ref. 1).

l 8.3.8 Station Blackout (Loss of AC) (FSAR Section 14.1.2.8)

1. Description All AC power to the unit is lost, with only battery power available. The reactor and turbine trip, and reactor coolant and feedwater pumps are lost. Core cooling is accomplished through heat rejection to the secondary side using the turbine driven emergency feedwater pump with steam relief to the atmosphere. The analysis is performed starting at full power 2535 Mw (t), and takes credit for a condensate inven-tory of 200,000 gallons. NNI and ICS instrumentation is taken credit for in controlling the plant when it is powered from the vital ac inverters.

2. Acceptance Criteria

1. DNBR is not less than 1.3.

1457 265 8-8 Am. 5

11. Reactor coolant system pressure does not exceed code allowable pressure of 2750 psig.

3. Mitigation

1. Control of the steam driven emergency feedwater by the EFW 1evel control system.

ii. Steam relief through the atmospheric dump and main steam relief valves either by the ICS or in accordance with Emergency procedure 1202-2 and 2a. g

4. Conclusion The FSAR analysis of this event remains bounding for the modified IMI-1 plant design. None of the plant ~odifications being made affect the systems and components which are necessary to mitigate this accident. Since the ICS is powered from the vital ac system, monitoring instrumentation will be powered by the station batteries. The operator will have all of the instrumentation available to bring the plant to a stable shutdown condition.

The extended analysis of this event will determine:

1. When power would have to be restored to maintain stable shutdown.

ii. RCS system pressure response without pressurizer heaters av ailable . 8.3.9 Steam Line Failure (FSAR Section 14.1.2.9)

1. Description A steam line rupture results in depressurization of the secondary system. This depressurization causes a primary system cooldown causing a DNBR transient and a positive reactivity addition. Blowdown can cause a significant mass and energy addition to containment. Finally, offsite doses can result from the release of secondary side steam to the atmosphere, if steam generator tube leakage exists. The FSAR analysis addresses a variety of break sizes, including the rupture of all four main stcam lines outside the reactor building. HPI was not assumed to operate during this event.

2. Acceptance Criteria

1. The core will be maintained in a coolable geometry.

ii. No steam generator tube loss of integrity will result from the pressure / temperature transient. iii. Offsite doses will be within the limits of 10CFR100. 1457 266 8-9 Am. 5

3. Mitigation

1. Reactor trip on low pressure or high neutron flux.

11. Feedwater isolation of the affected OTSG as a result of low steam generator pressure.

iii. Isolation of the unaffected steam generator by the turbine stop valves. iv . Decay heat removal through the unaffected OTSG by manual control of emergency feedwater (Procedure 2203-2.3) and either atmospheric dump valves or the turbine bypass valves if they are available.

v. Containment temperature and pressures are limited by the containment fan coolers (and reactor building spray systems if reactor building pressure exceeds 28 psig). l

4. Conclusion Recent, detailed analyses of TMI-2 (Ref s. 5 through 8) allow broader conclusions about the acceptability of TMI-1 regard-ing steam line break. The E41-2 analysis considered addi-tional single failures, the'most limiting were the feedwater regulating and turbine stop valve failures. In addition, the reactor core performance was analyzed assuming that: feed-water is not isolated, of fsite power is available if results are worse for that case, and both steam generators blow down outside containment. Reference 3 explains why the TMI-2 core performance analysis bounds Unit 1.

At the Cycle 5 refueling outage, the feedwater latching signal was added to the upstream block valves (FW-V-SA/B). The TMI-2 feedwater regulating valve and turbine stop valve failures cases thus bound the IMI-1 design. Although these failures are not a licensing basis for the plant, they do demonstrate the additional safety margins available in this accident. The difference in design of the main steam isolation valves between TMI-1 and TMI-2 results in less severe containment transients for TMI-1. The Unit I valves are a stop/ check design, so that they would prevent the blowdown of both steam generators inside containment. Since TMI-l does not have cavitating venturi's on the emergency feedwater lines, the operator would have to isolate the af fected steam generator to prevent containment overpressure. The operator would have approximately 20 minutes to perform this action. TMI-l has not analyzed the environmental ef fect inside containment for the worst case single failure (because of the 1457 267 8-10 A=. 5

stop/ check MSIV's, the worst failure is the feedwater regulat-ing valve f ailure). As noted previously, the blowdown will be less severe than for Unit 2. Although this issue is still being resolved, there are several reasons to expect accept-able results.

i. Heat shrink tubing is being added to splices inside containment. This change wr.s made to TMI-2 prior to receipt of the operating license to resolve this concern.

ii. Much of the equipment which was analyzed and shown acceptable for TMI-2 is also used on TMI-1. The radiological consequences of the unmitigated steam line break accident have also been addressed on the TMI-2 docket (Ref. 6 and 7). These analysis results demonstrate that worst case doses from a steam line break acc.' dent are within the limits of 10CFR100. 8.3.10 Steam Generator Tube Faila.e (FSAR Section 14.1.2.10)

1. Description The rupture of a steam generator tube concurrent with 1%

failed fuel results in the release of radioactive steam to the environment via the condenser air ejector. leakage is greater than the capacity of the makeup system, so that the RCS depressurizes.

2. Acceptance Criteria

1. Doser are less than 10CFR100 limits.

3. Mitigation

1. Reactor trips on low pressure.

ii. High pressure injection initiates and maintains primary system pressure and inventory. iii. Turbine trip isolates the steam generator, and the release path of steam to the environment is via the turbine bypass line, through the condenser to the air ejector. iv . Cooldown is achieved first via the unaffected steam generator and then through the decay heat cooling system. 4 Conclusions There have been no plant changes which change the results of this analysis. Results are still valid and acceptable. 1457 268 8-11 Am. 3

8.3.11 Fuel Handling Accident (FSAR Section 14.2.2.1 and References 8 through 10)

1. Description Failure of a spent fuel assembly, either in the fuel handling building or inside the containment building can result in release of radioactivity to the environment. The fuel handling accident in the fuel handling building considers a 72 hr. decay period for the fuel with the release of gap activity from the entire row of fuel pins on one assembly.

100% of the noble gases and 1% of this iodine inventory is released from spent fuel pool. The fuel handling accident inside containment assumed failure of an entire assembly, filtration by the refueling canal water, and release via the purge exhaust filtration system.

2. Acceptance Criteria

1. Doses should be appropriate within the guidelines of 10CFR100 (less than 100 REM).

3. Mitigation

1. Filtration of releases through the fuel handling building ventilation system.

ii. Filtration of releases by the purge exhaust filter system for the accident inside containment. iii. Meteorological dispersion of 6.8 x 10-4 sec/m3 for the accident initiating inside containment.

4. Conclusion The plant design changes do not affect the mitigation of the fuel handling accident inside containment. Results are still within the acceptance criteria.

The partitioning of the fuel handling building between Unit I and Unit 2 does not affect the consequences of this accident because each unit has its own HVAC system. A Unit 1 fuel handling accident would scill be mitigated by the Unit I ventilation system. 8.3.12 Rod Ejection Accident (FSAR Section 14.2.2.2)

1. Description Failure of a pressure barrier component could result in the rapid ejection of a control rod from the core. A power excursion and leakage sof radioactive primary system fluid to the secondary side would result. Releases to the environment result both from releases via the secordary system and from leakage from containment.

1457 269 8-12 Am. 3

2. Acceptance Criteria

1. The reactor coolant pressure boundary is not further degraded as a result of the ejected rod (no reactor vessel deformation).

ii. Offsite doses are within the limits or 10CFR100. iii. Radially averaged enthalpy should not be greater than 280 cal /gm at any axial location in any rod.

3. Mitigation

1. The power excursion is limited by the Doppler coef ficient.

ii. The power excursion is terminated by reactor trip on high pressure or high flux.

4. Conclusions The lower high pressure trip setpoint results in increased safety margins over the FSAR analysis. Improvements to the containment isolation signal (radiation +Rx trip) make release of fluid from the containment building less likely.

8.3.13 Feedwater Line Break Accident (TMI-2 FSAR, S3-22.49)

1. Description This event has not been analyzed for TMI-1. The following description is based on FSAR analyses for TMI-2. A loss of feedwater flow results in a loss of heat sink, primary system heatup, increased pressurizer level and pressure, and reactor trip on high RCS pressure. The TMI-2 analysis assumes a complete loss of feedwater due to a break upstream of the first feedwater line check valves. No analysis of loss of feedwater due to pump trip or valve closures were analyzed.

The loss of feedwater flow due to the postulated break is analyzed as an immediate loss of flow, which results in a bounding analysis for loss of feedwater events. The reactor is initially at 2772 Mw(t). Assumptions were made to provide two worst case scenarios one for containment, and one for primary system conditions. A double ended rupture (with a blowdown area limited by the feedwater nozzle area) was analyzed; steam generators are assumed to have a fouled inventory of 62,500 lbs. , and emergency feedwater is assumed to be at full flow within 40 seconds. The loss of feedwater is not directly calculated but taken as a conservative loss of heat demand (100-0% in 5 seconds for the affected generator and 100-0% in 20 seconds for the unaffected generator). 1457 270 8-13 Am. 3

2. Acceptance Criteria

1. Core thermal power shall not exceed 112% of rated power.

ii. Reactor coolant system pressure shall not exceed code allowable limits of 2750 psig.

3. Mitigation

1. Reactor coolant system trip on high pressure.

ii. The secondary system heat sink is restored by initiation of emergency feedwater to f ull flow within 40 seconds. Heat removal is through the turbine bypass valves or main steam relief valves.

4. Conclusions Results of the TMI-2 feedwater line break accident have become bounding for Unit I with the addition of a feedwater line break initiating signal. The addition of reactor trip or loss of feedwater increases the safety margin over the TMI-2 analysis. Lowering of the high pressure trip setpoint also increases safety margins since reactor trip will be initiated sooner. The RCS heatup is thus reduced. PORV operation was not assumed in the feed line break analysis, so that the increase in the valve setpoint does not af fect analysis results. The PORV would actuate for the worst case feedline break accident analyzed in the TMI-2 FSAR.

8.3.14 Waste Gas Decay Tank Rupture (FSAR Section 14.2.2.5)

1. Description The rupture of a waste gas decay tank would result in radio-logical releases via the plant ventilation system. The tank contents as calculated assuming the activity evolved from degassing the primary coolant system af ter operation with 1%

failed fuel.

2. Acceptance Criteria Doses shall not exceed the limits of 10CFR100.

3. Mitigation Elevated release of activity f rom the unit vent.

4 Conclusions This analysis has not _ been changed as a result of any plant modifications. 1457 27i 8-14 Am. 3

8.3.15 ymall Break Loss of Coolant Accidents ( .JCA)

1. Description Small break LOCA's are piping ru tures whose break areas range from as small as 0.005 ft. to as large as 0.5 ft.2, These LOCA's may or may not involve depressurization of the Reactor Coolant System (RCS).

2. Acceptance Criteria

1. Local fuel cladding oxidation (metal water reaction) sha.1 not exceed 0.17 times the total cladding thickness, or .05 the overall cladding mass.

ii. Paal Cladding Temperature (PCT) shall not exceed 2200*F. iii. A coolable geometry shall be maintained. iv . Long term cooling shall be assured.

3. Mitigation

1. Inventory will be maintained by the high pressure injec-tion system, ii. Emergency Feedwater flow within 20 minutes of very small break LOCA's allows depressurization of the RCS ,

and allows suf ficient inventory addition by the HPI system to maintain core cooling.

4. Conclusion Pursuant to NRC regulations (10CFR50.46) and 10CFR50 Appendix K) B&W performed generic LOCA analyses of their 177 fuel assembly lowered loop plants. Initially this work was per formed to meet the Interim Acceptance Criteria (IAC) and documented in BAW-10052. Later, the analyses were revised to the Final Acceptance Criteria (FAC) using the approved Appendix K model (BAW-10104). The FAC analysis results were documented in BAW-10103.

The work performed for BAW-10052 was used as the basis for the small break LOCA lcoation and size seasitivity study and therefore no new work was performed f r. BAW-10103 other than analysis of three specific break sizes and locations (0.04 ft.2, o,44 ft,2 and 0.5 ft.2 break sizes). In April 1978, B&W identified an error in their ECCS model. The error was also evident in the model used for the BAW-10052 sensitivity ctudies and therefore the basis for the accepta-bility of the small break analysis was eliminated. B&W performed additional small break studies using the corrected model. The revised analyses are documented in a letter f rom J. H. Taylor, B&W to S. A. Varga, NRC dated July 18, 1978. These analyses cover break sizes 0.04, .055, .07, .085, 0.1, 0.15, 0.2, 0.3, 0.13, and 0.17 ft.2, 1457 272 8-15 Am. 3

Key assumptions for the small break LOCA analyses versus the TMI-l plant design are given below: BAW-10103 Generic TMI-l Reactor Power (MWt) 2772 2335 Reactor Trip (psig) 1900 1900* RC Pumps (LOOP) Coas tdown Coas tdown AFW Available** Yes-40 sec. Yes**** ESFAS HPI (psig) 1600 1500 Operator Action Ye s-c ros s-connect none*** HPI Distribution 70% to Core 70% to core within 10 min. from time zero*** HPI Flow (gpm) 450 at 600 psig 500 at 600 psig

• Variable low pressure at full power.

  ** Amount assumed for generic analyses 550 gpm which is less than the minimum 900 gpm available for TMI-1. Results of Reference 2 demonstrate that EFW is not required before 20 minutes.
*** Prior to startup TMI-1 will install HPI injection leg cross connects and flow control devices to eliminate operator action to cross connect HPI and equalize flow in all four injection legs.

• • • • For worst case LOCA in which of fsite power is lost, EFW is initiated by the control grade loss of feedwater signal.

17 all cases, TMI-1 plant specific information is as conservative or more conservative than the generic r.ssumption. Since the TMI-2 accident, greater focus has been placed on small break LOCA's and the capability of the ECCS to mitigate them. Problems such as those discussed in Reference 21 (where the pressurizer stays full due to the loop seal arrangement despite loss of RCS inventory) have been addressed. These studies are documented in B&W's " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" May 7, 1979 (Reference 2). Breaks of 0.01, 0.02, and 0.07 ft.2 are analyzed utilizing varying assumptions on the availability and timing of AFW and HPI. These analyses use the same initial assumptions as used in BAW-10103 except that ESFAS is assumed to occur at 1350 psig. Therefore, they are also bounding assumptions for TMI-l except for the distribution of HPI flow as discussed below. The analysis in Reference 2 also established that EFW flow is not required less than 20 minutes before any steam line break accident. 1457 273 8-16 Am. 5

In Reference 2, credit is taken for operator action to initiate HPI or EFW. No mention is made as to whether whether operator action includes the time necessary to cross connect HPI as required in B&W's other small break accident analyses. TMI-1 will com-plete the installation of permanent cross connection of the HPI prior to star':up, therefore, operator action will not be neces-sary. All of the B&W small break LOCA analyses assume essen-tially equal backpressure for all four HPI injection points. This assumption is the basis for the 70%/30% flow split of HPI (assuming a single f ailure of one HPI train) between the core and the break respectively, af ter cross connection is accomplished. Such an equal backpressure would not exist given an HPI line rupture. The back pressure on the broken HPI leg would be essentially zeru and therefore the HPI loss out the break could be high resulting in inadequate injection to the core. The criterion established by B&W for the small break analysis requires that 70% of the total flow for one HPI pump be injected into the broken legs of the reactor coolant system. This cri-teria applies to a 2772 MW thermal 177 fuel-assembly plant. For TMI-1 with a licensed core power of 2535 MWt, the 70% - 30% criterion can be relaxed in direct proportion to the power reduction. This is justified based on the f act that the decay heat load following a small break LOCA is proportional to power and therefore cooling requirements will be directly proportional to the power at which the plant has operated. Therefore, for TMI-1, the acceptable flow split can be relaxed to 64% - 36%. The 64%/36% flow split would not be obtained for an HPI line break as explained above. Therefore, operator action would be required to isolate the ruptured HPI line. The need to isolate could be determined by observing the individual flow indicators for the HPI legs. The high flow leg would then be isolated. This action would be contrary to the operators instinct and would require considerable judgment since the initial flow imbalance may not be dramatic. Since too great a chance for operator error (error of omission) exists, cavitating venturis will be added to the injection legs to limit flow in the broken leg. The venturis have been sized to limit flow in each leg to 137.5 gpm when only one high pressure injection pump is cperating and Reactor Coolant System is at atmospheric pressure. The venturi design ensures that for the worst case HPI line break condition, the 64%/36% flow split can be achieved when Reactor System Pres-sure is less than 1500 psig. At RCS pressure conditions greater than 1500 psig, a flow split beyond the 64%/36% acceptance criteria will occur. B&W has reviewed this situation and judged the cavitating venturi performance is ac ?eptable. This conclu-sion is based on the fact that under HPI line break conditions, the Reactor Coolant System will not expend significant time above 1500 psig and that during the time the RCS is above 1500 psig the cavitating venturi ensures that there is significant flow of high pressure injection into the RC system. B&W also notes that a much larger small break than a HPI line break sets the generic flow split criteria and therefore for a HPI line break the flow split criteria can be relaxed. 8-17 Am. 2

In addition to the benefits discussed above, the venturis provide two added benefits. First, they balance flow of the injection legs under all other small break conditions such that TMI-l flow split will be within the bounds of the generic analysis (i.e. , 70*./3-% flow s plit) . Secondly, the cavitating can be relaxed. 8.3.16 Large Break Loss of Coolant Accidents (Reference FSAR Section 14.2.2.3)

1. Description Break sizes in the reactor coolant system (RCS) greater than 0.5 ft.2 are classified as large break loss of coolant accidents (LOCA's). These breaks involve rapid depressuri-zation of the RCS and are accompanied by rapid increases in containment pressure. Offsite doses are calculated from the design basis radioactivity release to containment, and the design basis containment leak rate.

2. Acceptance Criteria

1. Peak fuel clad temperature does not exceed 2200*F.

ii. The core is maintained in a coolable geometry. iii. Local fuel cladding oxidation (metal water reaction) shall not exceed 0.17 times the total cladding thick-ness of .05 times the total cladding mass. iv . Of fsite doses are within the li=its specified by 10 CFR 100.

3. Mitigation

i. Core flood tank actuation at 600 psig to establish water inv ento ry.

ii. Low pressure injection system flow below 200 psig to establish core cooling for the remainder of the accident. iii. Building spray addition to put iodine in solution with the containment water volume thus preventing release to the environment. iv . Containment leak tightness to limit radioactivity releases.

v. Switchover of the decay heat removal system suction source to the containment building sump on low-low BWST level.

4 Conclusion The calculated of fsite dose resulting from the design basis LOCA will increase as. a result of the deletion of sodium thiosulfate from the building spray system. Doses will still 1457 275 8-18 Am. 5

be within the limits of 10 CFR 100. Dose calculations per-formed for TMI-2 (see TMI-2 FSAR, Section 15 and Reference 5) demonstrate that design basis LOCA doses are within the limits of 10 CFR 100. The TMI-2 dose calculations were performed taking no credit for sodium thiosulfate. Since Unit 2 has a slightly large thermal power level and allowable containment leak rate, then Unit 2 dose calculations conservatively bound the worst case LOCA dose for TMI-1. Automated switchover of the BWST to the recirculation mode provides additional assurance that switchover will occur within the correct level band. Correct operator action had always been assumed in previous LOCA analyses. The automated switchover achieves the same function requirement by means of a safety grade control system. 8.4

SUMMARY

AND CONCLUSIONS Plant modifications to TMI-1 allow the plant analyses to bound the expected plant behavior (see below). In some cases, analysis for TMI-2 have been referenced because they either analyze events that are not in the TMI-1 FSAR (feedline break) or provide additional assurances of safety margins (steam line break).

1. Raising the PORV setpoint and lowering the high pressure trip setpoint af fects all of the pressurization transients in the FSAR. Safety margins are improved since the high pressure trip setpoint has been lowered. No credit was taken for operation of the PORV, so that raising tha valve setpoint has no ef fect on the FSAR analysis results.

The combined ef fect of the PORV and RPS setpoint changes are to decrease the probability of PORV operation. The integrity of the primary coolant system will be challeng-ed less frequently, so that this change is in the conserva-tive direction. It should be noted that this modification could result in more frequent plant trips.

2. Reactor trip resulting from loss of feedwater results in improved safety margins for loss of feedwater events and does not degrade plant response for any accidents / transients.

3. Reactor trip as a result of turbine trip increases safety margins for the loss of feedwater or feed line break analy -

ses. The effect of retaining or deleting plant features that permitted this event to occur without a reactor trip is being analyzed.

4. The adultion of emergency feedwater initiating signals for the feedline break accident makes the TMI-2 feedwater line break accident analysis bounding and conservative for TMI-1.

This event has additional safety margins beyond the TMI-2 analysis since both turbine and feedwater trips result in a 1457 276 8-19 Am. 5

reactor trip. This earlier reactor trip will result in a smaller heatup of the primary system.

5. Modifications to the high pressure injection system will al-low adequate HPI flow for the spectrum of LOCA's. Systeu per-formance is not degraded for any other accidents / transients in which HPI flow is initiated.

6. Upgrading of instrumentation inside containment assures that instrumentation will be functional in the postulated accident environments.

7. Automated switchover of the BWST to the recirculation mode provides additional assurance that switchover will occur within the correct level band. Correct operator action had always been assumed in previous LOCA analyses. The automated switchover achieves the same function requirement by means of a safety grade control system.

8. Dose calculation performed for TMI-2 demonstrate that the requirements of 10CFR100 are met even af ter sodium thiosul-fate is deleted.

9. Partitioning of the fuel handling building does not degrade the capability of the building HVAC to mitigate fuel handling accidents. Filtration of radioactivity will still be accom-plished in accordance with the licensing basis for the unit.

10. The transition to natural circulation following a c aplete loss of feedwater will be demonstrated by a startup test.

11. An analysis of the station blackout will be performed to determine what specific actions would be required to bring the plant to a safe shutdown condition.

12. A PORV setpoint of 2450 psig does not result in unacceptable interactions between the PORV and the pressurizer safety valves, whose setpoint is 2500 psig.

1457 277 8-20 Am. 5

REFERENCES

1. Three Mile Island Unit 1 Nuclear Station, Final Safety Analysis Report, USNRC Docket No. 50-289.

2. " Evaluation of Transient Behavior and Small Reacter Coolant System Breaks in the 177 Fuel Assembly Plant," Volumes I & II, Babcock and Wilcox, May 7, 1979.

3. "GPUSC Safety Evaluation Report for Three Mile Island Unit 1 Cycle 5 Reload," dated March 1979.

4. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid) on "High Pressure Trip and Pressurizer Code Safety Valve Settings," GQL-0069, April 17, 1978.

5. " Supplement No. 2 to the Safety Evaluation Report by the of fice of Nuclear Reactor Regulation, Three Mile Island Nuclear Station Unit No. 2, Docket Number 50-320," USNRC, NUREG 0107, dated February,1978.

6. Letter, Met-Ed 'J. G. Herbein) to USNRC (S. A. Varga), on " Analysis of Fuel Performance During a Steamline Break for "3I-2," License No.

CPPR-66, Docket No. 50-320, dated November 18, 1977.

7. Letter, Met-Ed (J. G. Herbein) to USNRC (S. A. Varga), on " Response to Staf f Questions on Analysis of Fuel Performance During a Steamline Break," dated December 9,1977.

8. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid) on "IMI-1 Fuel Handling Accident Inside Containment," GQL-0460, dated April 20, 1977.

9. Letter, USNRC (R. W. Reid) to Met-Ed (J. G. Herbein), dated February 4, 1979.

10. Let er, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid) on "IMI-l Fuel Handling Accident Inside Containment, "GQL-0460, dated May 8, 1979.

11. ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103, Rev. 2, Babcock & Wilcox, April 1976.

12. USNRC to Met-Ed " Order for Modification of License ," Docket No. 50-289, 1-ey 19, 1978.

13. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid), on "Small Break LOCA," GQL-0809, May 3, 1978.

14. Safety Evaluation and Environmental Impact Appraisal by the Of fice Nuclear Reactor Regulation, Supporting Amendment No. 65 to Facility Operating License No. DPR-47, Amendment No. 62 to Facility Operating License No. DPR-55 Duke Power Company, Oconee Nuclear Station, Units Nos.

1, ' and 3, Docke t Nos. 30-269, 50-270, and 50-287, October 23, 1978.

15. TMI-l Fuel Densification Report, BA'J-1389, Babcock & Wilcox, June 1973.

1457 278 8-21 Am. 3

REFERENCES - (Cont'd)

16. GPUSC Safety Evaluation Report of B&W's TMI-1 Cycle 4 Reload Report, dated January 13, 1978.

17. GPUSC Safety Evaluation of B&W's IMI-l Cycle 3 Reload Report, dated January 21, 1977.

18. Office of Standards Development, U.S. Nuclear Regulatory Commission, Regulatory Guide 1.70, " Standard format and Content Guide , Rev. 3, LWR Edition.

19. " Order and Notice of Hearing, Docket 50-289," dated August 9,1979, US NRC.

20. "IMI-2 Lessons Learned Task Force Status Report and Short-Term Recom-mendations, NUREG-0578, July 1979.

21. Michelson, C. "f ecay Heat Removal During a Very Small Break LOCA for a B&W 205-Fuel Assembly PWR*. January, 1978.

22. Winks, Robert W., Analysis of the Requirements for Loss-of-Electrical-Load Transient at TMI-1. Babcock & Wilcox. October 28, 1975.

23. Winks, Robert W., Final Report on Phase 1 (Turbine Trip Test) and Evaluation for Phase 2 Testing for the Loss of Electrical Load Capa-bility at TMI-1. Babcock and Wilcox, May 20, 1975.

1457 279 8-22 Am. 5

                            /.SCIDENTS/ TRANSIENTS CONSIDERED FOR RE-ANALYSIS TABLE 8.2-1 Don't Consider Depend on     Non-Safety       Failure of   Analyses Need Affected by        Operator      Equipment        Non-Safety    to be more Plant Changes        Action       _ Used           Equipment     Realistic Startup Accident               X                                                X D!!ution Accident              X                               X                X Cold Water                     X                                                X loss of Coolant Flow                                                            X             X Dropped Rod                                                    X Ioss of AC                     X                X              X loss of Elec. Load             X                               X                X Steam Iine Failure                                             X                X             X Steam Generator                                                X                X Tube Failure Fuel llandling Accident                                        X Rod Withdrawal at              X                                                X Power Rod Ejection Accident          X Small Break LOCA               X                               X                X             X EVENTS NOT ANALYZED IN FSAR EFW Inadvertent                X                               X                X             X Initiation Loss of Feedwater              X                               X Feed Line Break                X llPI 1.ine Break               X                X              X loss of Offsite Power          X                                                X             X LT1 N

N CD CD

• f I .

S-I e h i A' 1457 281 m -

v. .

9.0 "AS-BUILT" DRAWINGS

9.1 INTRODUCTION

This section contains plant "as-built" drawings as the plant existed prior to the installation of any of the modifications which are discussed in Sections 2.1.1, 2.1.2, and 2.1.3. 9.2 RESTART MODIFICATION ADDITIONAL ELECTRICAL LOADS Table 9.2-1 lists the additional electrical loads that have been assigned. As new loads are identified for the long term modifi-cations and electrical bus assignments are made, Table 9.2-1 will be revised accordingly. 9-1 Am. 4 1457 282

TABLE 9.2-1 Restart Modification Additional Loads Additional Load MCC or Items Title in KW Dist. Panel Bus Upgrade Decay Heat System Vent Valve A 0.7 DC Pool IE Vent Valve B 0.7 DC Pool IF Vibration Monitor 0.6 MCC IB Bus IS 2.1.1.1 Reactor Trip Neg. 2.1.1.6.3.1 Incore Thermocoup!"s Neg. 2.1.1.5 Containment Isolation less than 1 2.1.1.2 Valve Position Indication 0.7? PNI ATE Inverter IE Computer Less than 0.5 Inverter IE 2.1.1.2 Power Operated Relief Valve Position Indication Less than 1 2.1.1.4 H2 Recombiner A bus heater & blowers 45 MCC 1A IP B bus heater & blovars 45 MCC IB is space heater 0.8 PNL CT-E IS isolation valves

• 0.53 DC PNL lA isolation valves 0.53 DC PNL IB position indication 0.025 Swing DC PNL IM (PNLS 1A and IB) 2.1.1.7 Emergency Feedwater Less than 1 Changes to HPI System to Accommodate Small Break 0.05 Inverter IE LOCA 2.1.1,3 Pressurizer Heaters connected only on loss of

__. of f site power and no E.S. 43, Heaters to A System 126 KW IP (y, Ikaters to B System 126 KW IS N 2.1.1.6.3.3 Psat Ala rm less than 1 N CD u

TABLE 9.2-1 (Cont'd.) Restart Modification Additional Loads Page 2 Additional Load MCC or Items Title in KW Dist. Panel Bus 2.1.1.6 Reactor Coolant System Temp. tkg . Reactor Building Sump Water Level Less than 1 Reactor Building Cooling Fan Motors No Requirement Fire Protection Change later

                ?;OTES

1. Items marked "neg." indicates less than 100 watts load ard the source has not been determined.

2. Items marked "no require;nent" indicates no load or no additional load requirements.

3. Items marked "later" indicates that changes are unknown. This inforestion will be supplied when available.

M W N N CO b.

4 a

   .e 4

Ie O 1457 285

10.0 CROSS REFERENCE TO ORDER REC 0HMENDATIONS

10.1 INTRODUCTION

The August 9, 1972 Order and Notice of HearinF issued by the Commission listed numerous actions recommende2 by the Director of Nuclear Reactor Regulation (NRR). These recc amendations are listed in Section 10.2. The section of this ceport that covers the recommendations is referenced or a reap anse is given. A number of the recommendations require additional guidance or have been amended /modofied during the course of their development especially those related to I&E Bulletins 79-05A, 05B,a nd OSC and NUREG 0578. Met-Ed is therefore responding to these recom-mendations as they are currently understood. 10.2 SHORT-TERM RECOMMENDATIONS AND MET-ED RESPONSES Recommendation Response 1(a) Auxiliary Feedwater Upgrading Section 2.1.1.7 1(b) Auxilary Feedwater Operating Sections 3.1.1 Procedures and 3.1.4 1(c) Control Grade Reactor Trip on Section 2.1.1.1 Loss of Turbine /FW 1(d) Complete Analysis for Small Break Section 3.1 LOCA's and Revise Procedures 1(e) Retraining of all Reactor Ocation 6.0 Operators 2 I&E Bulletins IEB 79-05A Item 1 Not Applicable Item 2 See Section 10.3.1 Item 3 Section 3.1.1 Item 4 Sections 3.1.1 and 6.2 Item 5 Sections 3.1.2 and 3.1.3

    .,.          Item 6                                   Section 2.1.1.5 Item 7                                   Sections 3.1.2 and 3.1.3 1457 286 10-1

Recommendation Response Item 8 Section 11.2.1 Item 9 Section 2.1.1. 5. 3 Item 10 Section 3.1.3 Item 11 (La te r) Item 12 Section 3.1 Table 3.1-1 (AP 1044) IEB 79-05B Items 1 & 2 Sections 3.1.4 and 6.2 Item 3 Sections 11.2.3 and 8.1 (Based on BSW analyses submitted May 7, 1979) Item 4 Sections 3.1.1 and 2.1.1.1 Item 5 Section 2.1.1.1 Item 6 Section 3.1 Table 3.1-1 (AP 1044) Item 7 Section 11.2 IEB 79-05C Item 1 Section 3.1.1 Item 2 (La ter) Item 3 Section 3.1.1 Item 4 Sections 3.1.1 and 6.0 Item 5 See NUREG 0578 Item 2.1.9 below Item 1 (Long Term) Sections (later) and 8.2.4

3. Emergency Plan Upgrading Section 4.0

4. TMI-1/IMI-2 Radwaste Ventilation Section 7.2 and Sampling Separation

5. TMI-l Radwaste Management Capa- Section 7.3 bility

6. Organization and Resources Section 5.0 to be supple-mented separately 1457 287 10-2 Am. 3

Recommendation Response

7. Financial Qualifications To be Submitted Separately

8. TMI-2 Lessons Learned Recom-mendations - NUREG 0578 2.1.1 Section 2.1.1.3 2.1.2 Met-Ed will participate in the EPRI/NSAC program to conduct performance testing of PWR rclief and safety valves. We wil.1 verify that the program in applicable to TMI-1. It is understood that this program will be reviewed with the NRC prior to testing to ensure that the intent of NUREG-0578, Re-commendation 2.1.2, is satisfied.

We believe that substantive test data can be obtained by July, 1981. However, scheduling of the test facility, acquisition of valves to be tested , and the possibility of extensive retest-ing could recult in a longer schedule. 2.1.3.a Section 2.1.1.2 2.1.3.b Section 2.1.1.6 2.1.4 Section 2.1.1.5 2.1.5 Section 2.1.1.4 2.1.6 (La te r) 2.1.7 Section 2.1.1.7 2.1.8 (Later) 2.1.9 Sections 3.1.1, 6.0 8.1 and (Later) z.2.1.a Section (Later) 2.2.1.b Section 5.0 2.2.1.c Section (Later) 2.2.2 Section 4.0 2.2.3 Not Applicable until NRC Regulations are revised i457 288 10-3 Am. 3

10.3 SPECIFIC RESPONSES TO RECOMMENDATIONS 10.3.1 Response to IEB 79-05A, Item 2 TMI Unit Transient Review The review of previous TMI Unit 1 transients has been completed as requested by ISE Bulletin 79-05A, Item 2. An in-depth study of the reactor trips which were not an integral part of a Startup and Test procedure (TP 800 series) constituted the major portion of this review. Transients such as rod drops and turbine trips without an associated reactor trip were given a brief review although these transients were not similar to the Davis Besse Ev ent (Enclosure 2 of I&E Bulletin 79-05). There were no significant deviations from the expected perform-ance during these past transients however there were two cases where the system response differed from it's " normal" response to the initiating event / reactor trip. Pressuricer level exceeded 400 inches as indicated by the Control Room stric chart and the post trip review for Trip #11. Average reactor coclant tempera-ture following Trip #12 dropped to approximately 536*F with a corresponding low level in the pressurizer of 16 inchec. The data from both of these trips indicate that no safety limits were exceeded and that prompt post trip recovery followed in each case. Reactor Trip #11 was initiated by an RPS High Pressure trip as a result of a turbine trip caused by high bearing vibration on August 30, 1974. The high pressurizer level occurred during the attempted run back from 75% FP when the RC average coolant tempe rature rose from 578* to 606.5 *F. Actuation of the PORV in conjunction with the ICS runback kept the reactor on line while operator action to cutback feed flow caused the high rise in Tave. The RC expansion due to the temperature rise in the system resulted in the high level exceeding the nominal range of the pressurizer. The RCS Pressure transient caused by the turbine trip and ensuing action resulted in a variable pressure / temperature trip. Recovery of RCS pressure, temperature and pressurizer control quickly followed the reactor trip. In addition to the high pressurizer level during Trip #11, the RC Drain Tank rupture disc was ruptured. This was most likely due to the pre-trip leakage to RC Drain Tank during normal operation compounded by the high energy blowdown through th.. electromatic relief when the pressurizer was essentially solid. This is not considered to be a deviation from the expected performance since the rupture disc performed as designed to prevent overpressuriza-tion of the Drain Tank when the energy input exceeded the cooling capacity of the system. Reactor Trip #12 on March 30, 1975, from 100% FP was initiated by an RPS High Pressure Trip following a turbine trip caused by a temporary loss of 125 VDC power to the EHC. Pressurizer level 1,_, 1457 289

dropped approximately 219 inches and T ave dropped to 536*F following the trip; both of these related parameters exceeded their normal response to a turbine trip. The apparent cause was the failure of two main steam safety valves, MS-V21A and MS-V20B, to reseat. This allowed an initial drop in OTSG header pressure to 950 psi until the turbine bypass valves adjusted to compensate for this additional steam relief. Although this was a deviation from the expected performance it is not considered to be signifi-cant since no Limiting Conditions for Operation were violated, pressurizer level remained on scale, and the turbine bypass valves were more than adequate to contro'i header pressure. Summary of Corrective Action Since the above review did not identify any significant devia-tions from expected performance, no major corrective actions were undertaken. Minor corrective action such as rechecking of instrument setpoints were performed. Details concerning each transient that was reviewed and the specific corrective actions taken are included in Appendix 10A. Neither of these transients were reported as reportable occurrences. 1457 290 10-5 Am. 3

10.4 LONG-TERM NRR RECOMMENDATIONS AND MET-ED RESPONSES To be provided later. 1457 291 10-6

APPENDIX 10A 1457 292

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3CC 55?. -! ? 555,47 558.5 149 I Si- 5* 1457 293

b. Tc . e

• n S/M/7'l (9 1312 I. Events Leading Uo g the Trio The reactor was operating at 75% power with steady state conditions.

All ICS stations were in AUTO. Vibration of the Number 11 bearing on the main turbine increased rapidly to the trip point (8 mils.). . This caused a turbine trip. In the one minute time span immediately following the turbine trip, the seccndary safety valves lifted, the primary system dropped slightly and the feedwater flow was cut back (See Enclosure II). The loss of steam load and the decrease in feedwater flow caused the primary pres-sure to increase rapidly to the electromatic relief valve setpoint. While the electromatic was lifting to decrease the RC pressure, the feedwater flow was still being cut back drastically. Feedwater de and had baen taker into manual at some point and manually reduced, but it isn't clear on the data and charts exactly when this occured. It appears that feec.ia ar h d cut back automatically at first and then again manually. The rasit . as that Tave began increasing rapidly as the primary pressure was decreasing. The operator tried to manually increase feedwater r.ow t: correct his error but the reactor tripped on Variable Temp./ Press. be-fcre the correction could be made. II. Etecrated Control System Resconse b Id

                                                                  /a cd]g Before the reacter trip, all ICS stations were in AUT0.         The turbine tripped at 1310:51 and at 1310:57 a Unit Load Limited t;y Feedwater Flow alarm was received alcng with a Steam Gen. B on BTU Limit alarm.         At 1310:59 a Unit Load Limitec by Reactor alarn was received and the reactor began runr.ing itself back to 15% power as designed.       At the same time a Steam Gen. A or BTU Limit alcrm came in.      At 1311:00 the Feedwater Flow Limit cleared but came right back in at 1311:01.      The Steam Gen. A cn ETU Limi      cleared at 1457 294

1311:04 The Unit Load Limited by Feedwater Flow alarm did not clear until 30 seconds after the reactor trip. When the first Feedwater Flow alarm (1310:57) was received, tne feedwater flow began running itself back in AUTO. Shortly thereafter, the CR0 took Feedwater demand into MAN. and ran feedwater flow down even faster.

• ICS response post trip was per design o- -

o D D \ A 8 I L d J-n i e ua

                                                                        .J    ,_2 III. Ooerator Action Before the trip, the operators were observing control rod movement be-cause of a recent deboration. When the turbine tripped off they put their attention on keeping the reactor from tripping.      When the primary CR0 saw pressure decreasing due to electromatic relief lift he placed feed-water demand in manual and began reducing 'eedwater flow to stop the pressura drop, but he realized he had dropped it too much when he saw Tave increasing rapidly.      He began increasing feedwater flow, but the reactor tripped before he could recover the feedwater flow.

After the reactor trip, the operators carried out reactor trip and em-ergency prccedures. IV. .O_v.eral' Pla-t :estense Immediately following the turbine trip the reactor began to run back in power as designed. It reached about 25% power before the reactor tripped. The turbine trip caused the secondary safety valves to lift. They re-mained lifted approximately 30 minutes and 6 to 10 of them lifted. The less of steam load caused the primary system pressure to increase until the electrcmatic relief valve lifted. The lifting of this relief caused the reacter coolant drain tank rupture disc to fail. There is no indication cf abnormal operation of the electrcmatic relief valves. It opened for a: proximately 39 secor.ds and then reseatec. 1457 295

The turbine header pressure was 890 psi. before trip and hit a max-imum of 1055 psi. The OTSG "A" Level was 177 inches before the trip and went to a minimum of 28 inches. OTSG "B" was at 170 inches and dropped to about 10 inches. During the transient the charts show a maximum R.C. Pressurire Level of greater thaa 400 inch's and a minimum of 100 inches.. The R.C. Pressure (WR) hit a maximum of 2309 psia. and a minimum of 1691 psia. Tave reached maximum of 606.5 0 F and then cooled down to 5320F afcer the reactor trip. Finally, R.C. Pressure (NR) reached a maximum of 2295 psia. Tables and charts of these values can be found in the en-closures. 1 O U [l g/ . Recommenda.ivos

1. Have General Electric representatives evaluate the problem with

        #11 baaring.

2. Check setpoints on electromatic relief.

3. See actached letter from B&W. PORC should recommend GAI evaluate capacity of cocling system.

4. Inscrument Department looking at some method of starting recorders in fast steed based on input from plant parameters during trips.

4 1457 296

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                                                    %     X f\f ee        eAQ l _UnL ara n a m P e ia                        3/3c/75 & ooso CAtsz or ara c a TRIP                P cessu e e H i3h CAU5E OF T;R322 HIP L -u.         c$     lOSVD6 pcm.w Ic          6 NE.

ara = a P0hn AT TD2 OF TRIP 10 0  ?.F. P. cPr; ca AcIIcss/,temccks ms vala 4 ms-Vzc3 dici ne i resed p.crerlj. Shcied (nu- i',6 opened MU-i/ NA; cic5ect mu-V 3, 'Tec k ,m o u c $ cen kei e bypass vealces (bebe.). PCs? E P PZvE i THE TAVE 'Hi Pressuri::er R.C. SI"CE (D) ( F) Tg (F) Level Pressure 'H'IP (inches) (PSIG) (Secon:is)

 -30          5iS          l-Oo          55(a       235              914 L-        -

0 562.5 567 555 937 17c a Sc 549 550 64tc ISS s '1 D H 44 GO 5%E 643 532 i to 9 7 90 53b S4e 532 ac 17 e a tac 53 6.5 559 534 i (o r759 166 53'i 640 633 29 1352 19 0 5 41, 5' li SC 541 50 #452 290 545. 6 64 L, 546 90 ai33 1457 299

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h. I r a Il 3/30/75' G. Oct o I. Events Leadinc Uo to the Trip The unit was operating at full power with steady state conditions. The "A" start up block valve was closed and the "B" main feed pump was in hand. All other ICS stations were in AUTO. There was no indication of any problems. At 0010 on March 30, 1975, the turbine tripped. The ,

operator received approximately ten overhead SCAM alarms and then checked the control rod pI panel and noted in limit lights for all control rod groups with the exception of group S, indicating the reactor had tripped. The operator then began trying to stabilize the plant. This all took place in a five to ten second time span. II. Integrated Control System Resoonse Before the reactor trip all ICS stations, except "B" main feed pump, were in AUTO. An erroneous signal from a faulty 701 relay indicated a loss of 125 vol DC supply to the turbine EHC system. This immediately resulted

    - in a turbine trip. In the next four seconds, the following events occurred:

ICS went into tracking mode, reactor power began decreasing, Tg began de-creasing, Tcbegan increasing, RC pressure increased, pressurizer level increased, feed water ficw decreased. Four seconds after the turbine tripced, the RC pressure reached 2355 psi and the reactor tripped. To this coint in time the control room operators took no action. RC pressure began to decrease rapidly and the pressurizer level started dropping. The control rec: o;erators began to take manual control of the piant in order to stabilize conditions,

                                                                 ,9%

III. Ocerator Actier Dm}D oe }ObN@g[ 3 0 $} The primary system cperator was at the console when the overhead alarms began annunciatir.g. He looked up and saw about 10 alarms flashing. One he noticed was high RC pressure. He then noticed the in limit lights for all centrol red groups except grouc S. By this time , RC nressure and pressuricer level - u ere decreasing, so the operator closed '"' '/3,S7e#I00

MU-V153 and started MUP-1A. Pressure continued to decrease so the operator opened MU-V14A, thereby adding borated water to the RCS from the BWST. The secondary CR0 checked that all generator breakers were open and started all turbine-generatcr oil pumps. He then checked the header pressure which was reading about 1000 psi. The secondary operator took manual control of steam bypass and FW-V15A/B and lowered turbine header pressure to 900 psi. He maintained the OT53 levels at thi-ty inches. When ccnditicns had stabilized, the Diamond control station was reset, FM-V16/Aa3 were placed in AUTO, steam bypass was placed in AUTO, MU-V14A and MU-V153 viere cicsed and v.U-Pl A was stopped. The above actions indicate the operators correctly carried out the necessary emergency procedures. IV. Overall Plant Rescense Under these circumstances, the turbine trip was inevitable. The plant followed the same sequence of events that took place during the loss of lead test frcm 105 which was performed during startup testing. The reactor tripped four secends af ter the turbine trip due to high RC pressure. Most of the secondary system steam safety valves lifted and two remained open for w roximatel s :ne hocr. After the reactor tripped due to high pressure, the RC cressure began decreasing rapidly and pressurizer level dropped. As described cre/icusly, the CRO's took the necessary actions to stabilize these con di tions . cIlcwing short form (<24 hrs.) precritical checks, the reactor was taken critical 5.7 hours after the trip and the turbine was put back on line about ei;h: hours later. Encicsure i provides the computer sequence of events printout. Enclosure 2 provides the alarn printouts. The change of plant parameters with time during the transient can be found in Enclosures 3-5. 1457 301

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