ML19305D662

Amend 16 to Restart Rept, Providing Addl Info Re Plant Mods

ML19305D662
Person / Time
Site:	Three Mile Island
Issue date:	04/11/1980
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19305D659	List: ML19305D658 ML19305D662
References
NUDOCS 8004150351
Download: ML19305D662 (54)

Text

. .-_ - . _ _ _ _

TABLE OF CONTENTS O

%) y 1.0 INTRODUCTIG AND REPORT ORGANIZATION 1-1 1.1 Introduction 1-1 i l

1.2 Report Organization 1-1 1.3 Abbreviations 1-2 7 i

1.4 Definitions 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 on Loss of Feedwater/ 2.1-1 Turbine Trip 2.1.1.2 Position Indication for PORV and Safety _ 2.1-3 3 Valves Q 2.1.1.3 Emergency Power Supply Requirements for Pressurizer Heaters, PORV, Block 2.1-5  ;

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 l 2.1.1.6 Instrumentation to Detect Inadequate 2.1-17 Core Cooling j 1

2.1.1.7 Auxiliary Feedwater Modifications 2.1-20  !

i 2.1.1.8 Leak Reduction Program For Systems 2.1-29a  !

Outside Containment l

l 2.1.2 Long-Term Modifications 2.1-30 i ,

2.1.2.1 Post Accident Monitoring 2.1-30 I I

l 2.1.2.2 RCS Venting 2.1-31 l l

\

l 2.1.2.3 Plant Shielding Review 2.1-31 l l  !

2.1.2.4 Post Accident Sampling Capability 2.1-33 l O

V I

2.1.2.5 Reactor Coolant Pump Trip on HPI 2.1-34 l

i 3o o4T5@3G x

I generated when any of the valves open. Calculations have been O made for saturated, liquid and two phase flow. A summary of these calculations is provided in Appendix 2A. Tests run by B&W i

i on the electromatic relief valve under reduced flow condiitons  !

have confirmed the validity of this approach. Because of the j straight-forward and well known relationships that exist between i flow conditions and differential pressure across the elbow, the  :

signal from one differen ial pressure transmitter can be confi-  !

dently predicted for any flow conditions. For this reason it has been concluded that operating tects, which would be difficult i

since they involve opening the PORV and relief valves, will not  ;

be required. l 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 i piping of pressurizer code safety valves RC-RV1A and RC-RV1B and electromatic relief valve RC-RV2. This piping is classified as N2, Seismic I. Analysis has been performed to demonstrate that li this modification will not degrade the integrity of the existing l Os pipe. The pipe classification has been maintained up to and )

including the instrument root valves. T.he 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. l All of the equipment inside containment for Pressurizer PORV and l safety valve detection will be seismically and environmentally i qualified. Work is underway to upgrade the portion of the system i outside containment. This involves specifying and procuring of additional equipment. This should be installed by January 1981.

2.1.1.2.6 Ins trumentation The output signals from the three differential pressure trans-mitters will be displayed on indicators in the control rocm. l 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 l computer for logging, trending, and alarm functions. j 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 {

i 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. l 2.1-4 -_ _

Am. 16 _ ,_

2.1.2 Long Tarm Modification 2.1.2.1 Post Accident Monitoring O 2.1.2.1.1 system oescription  !

Certain post accident monitoring capability will be provided in compliance with Reg. Guide 1.97, Rev.2 as discussed below. Pending the availability of appropriately qualified instrumentation and equipment, by January 1, the1981.

following modifications will therefore be completed 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. Redundant indication of pressure will be provided. .

Containment indication Water shall Level - Continuous be provided containment in the control room. A safety water level grade j wide range indicator from the bottom of containment to a level of '

10 feet Reg. Guide will1.97.

be installed in accordance with the requirements of 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 - Two safety grade containment radiation monitors that are physically separated  :

t shall be provided with recording display and continuous indicator presentation in the control room. The range of this monitor shall be 107 R/hr and shall detect photon radiation down to 60 Kev.

The design of the radiation monitors shall be provided in accordance with Reg. Guide 1.97 Rev. 2 (Dec. 1979). To our knowledge, manufacture of appropriately qualified equipment to satisfy these requirements will commence by July 1980. They  !

will be installed and operational by January 1,1981.

i High Range Effluent Monitor - One high range effluent monitors  ;

intended as the Long-Term modification shall be oeprational for each normal gas release point by January 1, 1981.

The range of these monitors shall be as follows:

I o Undiluted Containment Exhaust . .. ... . . . 105 uci/cc o Main Steem Lines . l

. . . . . . . . . . . . . . 102 uci/cc o Auxiliary & Fuel Handling Building Exhaust .. . 10 uci/cc 3

o Condenser OFF GAS Exhaust . .. . . ... . . .

105 uci/cc l

2-1-30 Am. 16

_. - . .. . - . . . - --~

Regulatory Guids 1.97 Rev. 2 (Dec.1979) will be followed for the design of high range effluent monitors. Vital bus power shall be employed for each system's modular assembly with

() the normal power supplying the monitor pumps with diesel gener-ations as back ups. Further descriptions of increased range capabilities are provided in Section 2.1.2.

High Range Effluent Radio Iodine & Particulate Sampling Analysis -

The existing sampling system will be expanded and will include the addition 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 ,

to iodine ratio. Counting of the cartridges will be by use of NaI crystal connected to a single or dual channel analyzer with appropriate window and discrimination settings for th 364 Kev gamma of I-131, or by use of a GELI/MCA syscem. The expanded  !

portion of the sampling system would be placed in service follow- t ing an accident and will be located in an applicable area exhibit-ing low background. The system will be on site and operable by 1 January, 1981.

Prior to incorporation of the expanded sampling system, procedures will be developed for the use of silver zeolite cartridges and ,

normal particulated filters for sampling with a Nal detector and a single or dual channel analyzer for iodine and gross particulate j release rate determination. Specific details to insure exposures  ;

j are maintained as low as reasonably achievable will be incorporated ,

into the procedures.

() These procedures will be available for NRC review prior to i

l restart or 1 October 1980 whichever occurs first.  !

l 2.1.2.2 RCS Venting I 2.1.2.2.1 System Description I Power operated 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 from the top of the pressurizer and the top of both hot l legs using existing connections on the reactor coolant piping. '

The discharge from the hot leg vents will be directed to the con-tainment atmosphere. The system is shown schematically in Figure 2.1.-11. '

The vents to the containment atmosphere will tie into existing [

hot leg vent piping outside the secondary shield wall. As part [

, of this modification, remote operation of the vent valves in the l existing vent line from the pressurizer to the reactor coolant j drain tank will be provided and the system will retain the exist-ing venting capability. Control and position indication for the ,

power operated vent valves will be provided in the control room. '

O 2.1-31 Am. 16 l i

The key parameters to be monitored with on-line instrumentation  ;

include containment hydrogen concentration, reactor coolant boron

() concentration and letdown failed fuel monitors. The on-line hydrogen monitoring capability has been previously described in this restart report. l An on-line boronmeter will be installed.

The conceptual design and schedule for Jnstallation will be forwarded to the NRC by March 1,1980. The existing reactor 8 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 t i

containment atmosphere sampling systems shall be performed. l Modifications shall be completed as necessary to ensure that personnel can obtain samples under accident conditions without inturring a radiation exposure to any individual in excess of 3 }

reme 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 {

i analyzed within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period previously mentioned. The chemical analyses to be considered shall include both boron and chloride analysis. The conceptual design for required modifi-  ;

l

() cations will be forwarded to the NRC by June 1,1980.

l r

RCS sampling can be accomplished within one hour and analyzsed in l

an additional hour under normal circumstances. Met-Ed's experience i sampling under accident conditions at TMI-2 has taught us that slow j 4 deliberate steps are necessary to prevent personnel overexposures.

Since an early indication of significant fuel failure is obtained

+

i from the failed fuel monitor and an on line baronometer is being (

provided it is considered unnecessary and imprudent to attempt to i draw a confirmatory sample in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. All necessary and appropriate operator and emergency actions can be taken based on the early indications.

In addition chloride analysis provides inform- f ation which is only useful in the long term and therefore is not immediately necessary. i 2.1.2.5 Reactor Coolant Pump Trip on RPI i

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 i conditions associated with a LOCA have been detected. This will  ;

i be accomplished by requiring that RCP trip be initiated when the i

Engineered Safeguards System has actuated Safety Injection and an increasing RC void fraction has been detected as indicated  ;

by low RC pump motor current. The proposed logic will preclude j

(} RC pump trip during those events such as severe overcooling or  !

j very small breaks where maintenance of forced cooling is very i  !

l <

2.1-39 Am. 16

/ \

V V P.sger i of 2 THMEE Mil 2 ISLAND UNIT No. I TABLE 2. M .

LIST OF CONTAINMENT ISOLATION VALVES REQUIRINC MfM)IFICATIONS Valve Line Method Normal Post Actual Valve Penetratten Valve Valve Stae, of Valve Position No. Service System Tag No. Accident Position _ Acti,ation Slanal Source Tg In. Actuation Positlen Existina Modified Indication Existina Modified Notes 108 Coateinemat Air RM CH-VI Sall 1 Air Opea Closed Closed Yes 1,80 8,22 6,10 Sample CM-V2 Sall i Air Opea closed Clemed Yes Cit-V 3 Sall I Air Open Closed Closed Yes Ol-W4 Ball i Air Open Closed Closed Yes 283 Stena Generater CA CA-V4A Clobe 3/8 EMO Open Closed Closed Yes I,10 1,4,5,6,10 Sample CA-VSA Clobe 3/8 Air open closed Closed Yes 214 Steam Generator CA CA-V4B Clobe 3/8 EMO Open Closed Closed Yes 1,10 1,4,5,6,10 Sample CA-VSS Clobe 3/8 Open Closed Air Closed Yes 302 Intermediate IC IC-V2 Cate 6 EMO Open Closed Open/ Closed Cooling Yee 8.LIO 3,,7,8,9,10 IC-V3 Cate 6 Air Open Closed Open/ Closed Yes Water Outlet Line 307 Dealm. Water to CA CA-Vl89 Cate open 2 Air Closed closed Yes 3.10 8,5,80 Reactor BullJang 309 LetJoun Line to les MU-W2A Clube 2-1/2 EMO Open Closed Closed Yes 1,80 1,4,5,6,10 Purification HU-V28 Clobe , 2-l/2 EMO Open Closed Closed Yes 1,80 1,4,5,6,10 D=minesalisers MU-V3 Cate 2-1/2 Air open Closed Closed 1,80 Yes 1,6,10 323 BC Makeup MU HU-VI8 Care 2-1/2 Air Open Closed Closed Yes 1,10 1,2,80 328 Pressuriser and CA CA-VI clobe 3/8 EMO Closed Closed Closed Yes 3,10 1,4,5,6,10 Reactor Coolant CA-V2 Cate 3/8 Air Closed Closed Closed Yes Sample Linee CA-V 3 Clobe 3/8 EHQ Closed closed Closed Yes CA-Vl3 Clobe 3/8 EMO Closed Closed Closed Yee 329 Reactor Coolant MU MU-V25 Clobe 4 EMO Opea closed Open/ Closed Yes 1,7,10 Pump Seal Retura - 3,7,8,80 HU-V26 Cate 4 Ai r Open Closed Open/ Closed Yes 330 Reactor Coolant WDC WDC-V3 Clobe 2 opea EMO closed Closed Yes 1,80 1,4,5,10 Drale Tank WDC-V4 Cate 2 Air Opea closed Closed Yes Vent 334 Reactor Coolant WDL WDL-V303 Cate 4 EMO Closed Closed closed Yes I,10 I,4,5,10 Drale Tank Pump WDL-V304 Cate 4 Air Closed Closed Closed Yes Discharge t 333 Intermediate IC IC-V4 Cate 6 Air Opee Closed Open/ Closed Yes Cuo!!ag Water I,LIO 3,7,8,9,50 g Supply Line

, 334 Intermediate IC IC-V6 Cate 3 Air Open closed closed Yes

• Coollag to I,LIO 3,7,8,9,50 i 6

CRDM Coolin8 e

Colts f e

+

b

- . . . -- - ~ . . - . . . . . . .- . . --. - . - - - --_

O O O Pg. 2 of 2 THNLE Pi1LE ISLAND UNIT NO. l .

TABLE 2.5-2 (CONT'D)

LIST OF CONTAINHENT ISOLATION VALVES REQUIRING MODIFICATIONS =

Valve Line He t tet braal Post Actual Valve Penetrition Valve Valve Site, of Valve Accident Position Position Actuation Sianal Source No. Service System Tag _NA Ty E In. Actuation Positten Estetina Hodified Indication Existina Hodified Notes 336 Reactor au!! ding AN AH-VIA Butter- 45 Air Closed Closed Closed Yes 4,80 4,4,5,10 Outlet Purge fly Line AH-VIS Butter- 48 EHO Closed closed Closed Tee fly 346 Reactor Coolant JtS NS-VI5 Cate 8 EMO Open Closed open/ Closed Yes 1,80 7,8,9,10,

• Piasp beor Cooling Water Supply 347 Reactor Centent NS NS-V4 Cate 8 EHO Open Closed Open/ Closed Yes 3.10 7,8,9,30 Pump mtor NS-v35 Cate 8 Closed Cooling Water EHO Open open/ Closed Yes M 7,8.9.10 Return 353 Reactor sui! Jing WDL Wut-V534 Cate 6 Air Closed Closed Closed Yes 1,80 1,4,5,80 Sump Drain WDL-v535 Cate 6 Air Closed Closed Closed Yes 421 Reactor Building R8 RS-V2A Cate S ENG g Closed open Yes 1,10 Normal Air Ql0; AJJ auto initiation j Coolere Supply { ymg. R.S. coolina on 4 pela Line R.S. and 1600 pela B.C.

422 Reactor Building s mure isolation e5anale.

RS RS-V7 Cate 8 Air Open Closed Opea Yes Qlo; Normal Air 3.10 Coolers Reture Line 423 Reactor Building AN AH-VIC Butter- 48 Closed EHO Closed Closed Yes 1,4,10 1,4,5,10 Inlet Purge fly Line AH-VID Butter- 48 Air Closed Closed Closed Yes fly 348,349 Core Flood TE. CF CF-V2A&4 Clobe END Closed Closed Closed Sample and N2 Fill -VISA &B Cate 1

I Air Yes M I,5,80 1.ine s -V20A&B Cate  ! Air V-Ive Actuation Slanal Source

1) 4 pais reactor building preneure isolation 7) Claaetty !!as to Salaale Category 1

2) 1600 pais (SFAS) laulation 8) 30 pois reactor buildtes pressure teolation

3) Rad!1 tion alarm, operator action required 9) Line break isolattom signal

4) Migh radiation (noo-safety) isolattoa 10) Remote manual control y 5) Reactor trip isolation f*

6) Ovirride capability on individual valves 3

i I

t THREE MILg ISLAND UNIT NO. !

TABLE 2.l;3 LIST OF CONTAINMENT PENETRATIONS REQUIRINC ISOLATION ON HI-RADIATION Isolation Radiation

, Penetration Valve Detector Type of No. Service System Tag No. Location Konitor 213 Steam Cenerator CA CA-V4A Locate tie monitors outside tie R.B. Area and Sample -V5A near t he esspling line downstream of Camma 214 -V4B t he containment isolation valve and Detectors

-V5B upstream of connection for Turb. (New)

Plant sampling 309 Letdown Line to HU MU-V2A Utilise existing Rad. honitor RH/L-! Inline Purification -V2B located outside R.R. (Existing)

Demineralizers 328 Pressurizer and CA CA-VI Locate t he monitor outside t he R.B. Area Camma Reactor Coolant -V2 between the isolation valve and the Detector Sample Lines -V3 sample cooler. (New)

-Vl3 329 Reactor Coolant HU MU-V33A Locate t he online radiation monitor Area Camma Pumps Seal -335 downstream of t he containment isola- Detector Return -33C tion valves outside of t!e R. B. for (New)

-33D Alarm Operator action is required to close valves.

330 Reactor Coolant WDC WDG-V3 Locate t he monitor on t he outside of Area Camma Drain Tank t he tank. Detector -

and Vent -V4 (Existing) 331 Reactor Coolant WDL WDL-V303 Drain Tank , ,

Pump Disclerge -V304 336 Reactor Building AH AH-VIA Ut111:e the existing purge outlet Inline Outlet and -VIB line Rad. Monitor RM/A-9 located (Existing) and Inlet Purge -VIC outside of R.B.

423 Lines -VID 353 Reactor Building WDL WDL-V 534 Locate an area radiation monitor Sump Area Sump Drain -V535 in t he R.B. Sump mounted inside Monitor a seismically supported pipe. (New) g 302 Intermediate Cooling IC IC-V2,3 Locate t he radiation monitor on the Strap 333 Supply & Return -V4,6 6" IC return line between valve on GM

• and IC-V3 and t he 2" pump recirc. line.

=

334 (New) e o

O O O .

THREE MILE ISLAND UNIT NO. 1 -

TABLE 3

^

List of Isolation Signal Override Capability Isolation Signal Penet ra tion Reactor High 4 psig 30 psig 1600 psig Line No. Trip Radiation Building Building (SFAS) Break Containment Air Sample 108 N/A N/A C N/A C N/A R.B. Sump 353 C IB C N/A N/A N/A RCDT 330,331 C IB C N/A N/A N/A RCS Sample 328 C IB C N/A N/A N/A R.B. Purge 336,423 C NO NO N/A N/A N/A N

RCS Letdown 309 A IB C N/A N/A N/A RCS Makeup 323 N/A N/A C N/A C N/A Demin Water 307 C N/A C N/A N/A N/A 1

OTSG Sample 213, 214 C IB C N/A N/A N/A NSCCW 346, 347 N/A N/A N/A NO N/A NO ICCW 302, 333, N/A N/A N/A NO N/A NO 334 R.B. Air Coolers 421, 422 N/A N/A C N/A C N/A Core Flood TK 348,348 C N/A C N/A N/A N/A Legend C = Common Signal Override; initiating isolation condition nay still exist.

I = Individual isolation signal override capability; procedures governing override to be developed.

IB = Individual isolation signal bypass capability A = Automatic isolation signal override.

h NO = No override or bypass capability; initiating condition must clear to allow reopening of valve.

N/A = Not applicable.

5 Note: For combinations of initiating signals taat are allowable, refer to Table 2.1-2.

j and is trucked offsite for disposal at a municipal sewage treatmect facility. In the event that any TMI-2 sanitary ,

facilities become contaminated with radioactivity, the ability l V to remove TMI-01 sanitary waste from the site would not be l reduced.  !

. l 7.2.6 Radiation Protection and Decontamination Areas TMI-l has, installed as permanent plant equipment, all fucilities j t

necessary to support the radiation protection and decontamination activities of a single unit. Any facilities in TMI-1, that in l the past had been shared by both units, will be used exclusively l by TMI-l subsequent to startup. Since the accident, TMI-2 has l established a separate radiation protection organization utilizing l dedicated personnel and equipment. External f acilities have been ,

established under the recovery program to replace facilities

previously shared with TMI-1. '

I 7.2.7 Nuclear Sampling and Radiochemistry Laboratory l

The Temporary Sample Sink System consists of tubing valves and l other equipment necessary to satisfy all Unit II sampling i requirements without utilizing the Unit I primary sample lab. l Tubing and valves will be provided to f acilitate sampling the  !

following; Reactor Coolant Bleed Tanks, Miscellaneous Waste i Holdup Tank, Mini Decay Heat System, Pressurizer Steam Space, Pressurizer Water Space, Reactor Coolant Letdown and the Fuel O

i Pool Waste Storage System.  !

The system will include all necessary tubing and equipment to l allow adequate sample recirculation from the sample source back to that source. In addition, any purge required into the Sample i Sink will be contained 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. l l

The system will also provide an in line Boronometer for measuring '

boron concentration in the samples drawn from the Unit II RCS. '

The Boronometer will be equipr.d with its own calibration equip- l 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 Nuclear Sampling Facility is scheduled to be opera- [

tional by June 1, 1980.

7.2.8 Industrial Waste Treatment Facilities i The Industrial Waste Filter System (IWS) and Industrial Waste Treatment System (IWS) have been installed to comply with the i

7-5 Am.16 l l

- , . - - _. -.- - ,.n-- - . , , . . , , , - - - - - + - .e -- - -n~.-

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 Liquid Radwaste System. In the event of a plant accident 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. Intermediate 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.

7.3.5.3 Improsed In-Plant Radioiodine Monitoring Instrumentation See section 2.1.2.1.1.

O ~.4 Affect of TMI-2 Recovery on TMI-1 Operation Activities in TMI-1 related to radwaste processing and activities in the common fuel handling building will not be affected by the TMI-2 recovery program. As demonstrated in Section 7.2.3, TMI-1 does not have to rely on any T:tI-2 facilities for the processing of radwaste. Section 7.2.1 descr.ibes specifics to be taken to isolate the radwaste piping systems of the two units.

Through the isolation of the piping system, interf ace between the two unit's radwaste systems will be eliminated. Waste processing activities related to TMI-2 will be performed in the fuel handling building during the recovery program. These activities will not affect activities in the auxiliary building because the areas will be separated by an environmental barrier (Section 7.2.2). Communi-cation of the air spaces (TMI-1/TMI-2) of the fuel handling building will be minimized with appropriate modifications of the ventilation equipment in the building. Continuous access to the fuel handling building is not required for the safe operation of TMI-1 (with environmental barrier installed).

( 7-15 Am. 16 l

l .

i 8.3.7 Loss of Electric Power (FSAR Section 14.1.2.8)

O 1. Description l f Separation of the unit from the transmission network results ,

in the trip of the turbine. A more severe transient occurs i if the ICS does not run back the reactor load demand. The i result is reactor trip on high pressure. Cooldown is accomplished through the atmospheric dump or steam relief .

valves. In the presence of failed fuel and primary to i secondary leaks, this event can lead to low levels of  ;

radioactivity release.

i

2. Acceptance Criteria l

1. DNBR shall not be less than 1.3.  !

i

11. Reactor coolant system pressure will not exceed code l allowable limits of 2750 psig.

3. Mitigation i

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).

• 8.3.8 Station Blackout (Loss of AC) (FSAR Section 14.1.2.8)

1. Description f

All AC power to the unit is lost, with only battery power availa ble. 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

i. DNBR is not less than 1.3.

O 8-7 Am. 16

t

3. Mitigation

() 1. Reactor trip on low pressure or high neutron flux.

s f

11. Feedwater isolation of the affected OTSG as a result of low steam generator pressure.  ;

iii. Isolation of the unaffected steam generator by the i 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. I Containment temperature and pressures are limited by the v.

containment fan coolers (and reactor building spray systems if reactor building pressure exceeds 28 psig).

4. Conclusion ,

P Recent, detailed analyses of TMI-2 (Ref s. 5 through 8) allow broader conclusions about the acceptability of TMI-l regard-ing steam line break. The TMI-2 analysis considered addi-tional single failures, the most limiting were the feedwater regulating and turbine stop valve f ailures. In addition, the reactor core performance was analyzed assuming that: feed-r O 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. '

s At the Cycle 5 refueling outage, the feedwater latching signal was added to the upstream block valves (FW-V-5A/B).  ;

The TMI-2 feedwater regulating valve and turbine stop valve ,-

failures cases thus bound the TMI-l design. Although these i failures are not a licensing basis for the plant, they do  !

demonstrate the additional safety margins available in this accident. l t

The difference in design of the main steam isolation valves between THI-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. The TMI-1 design isolates ,

emergency feedwater on low steam generator pressure (600 psig).

The TMI-2 containment analyses bounds since it was performed

)

assuming 1250 gpm EFW flow beginning at 2 see and continuing '

throughout the accident.

L TMI-l has not analyzed the environmental effect inside l containment for the worst case single f ailure (because of the

()  !

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8-9 Am. 16 )

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stop/ check MSIV's, the worst failure is the feedwater regulat- ,

ing valve failure). As noted previously, the blowdown will be

(} less severe than for Unit 2. This issue will be resolved in

(/ response to I&E Bulletins 80-01 and 79-22. There are several reasons to expect acceptable results.

1. Heat shrink tubing is being added to splices inside containment. This change was made to TMI-2 prior to j 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 i 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 acd dent are within the limits of 10CFR100.

8.3.10 Steam Generator Tube Failure (FSAR Section 14.1.2.10) ,

P

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. Accentance Criteria o

1. Eoses are less than 10CFR100 limits. ,

3. Mitigstion i

1. Reactor trips on low pressure. .

11. High pressure injection initiates and maintains primary  ;

system pressure and inventory.  !

iii. Turbine trip isolates the steam generater, 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 unaf fected 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.

O V

8-10 Am.16

REFERENCES O 1. Three Mile Island Unit 1 Nuclear Station, Final Safety Analysis Report, USNRC Docket No. 50-289.

2. " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks ,

in the 177 Fuel Assembly Plant, " Volumes I & II, Babcock and Wilco; May 7, 1979. '

3. "GPUSC Safety Evaluation Report for Three Mile Island Unit 1 Cycle 5 Reload, " dated March 1979.

I t

4. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid) on "High Pressure l Trip and Pressurizer Code Safety Valve Settings, " GQL-0669, April 17, 1978.

5. " Supplement No. 2 to the Safety Evaluation Report by the office of I 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 TMI-2," License No.

CPPR-66, Docket No. 50-320, dated November 18, 1977.

7. Letter, Met-Ed (R. C. Arnold) to USNRC (S. A. Varga), on " Response to Staff 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 "TMI-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. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid) on "TMI-l Fuel t Handling Accident Inside Containment, "GQL-0460, dated May 8, 1979.

i 11.

ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103, Rev. 2,  !

Babcock & Wilco; April 1976.

I 12.

USNRC to Met-Ed " Order for Modification of License," Docket No. 50-289, f May 19, 1978.

l

13. Letter, Met-Ed (J. G. Herbein) to USNRC (R. W. Reid), on "Small Break  !

LOCA," GQL-0809, May 3, 1978.  !

14. Saf ety Evaluation and Environmental Impact Appraisal by the Office 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, 2 and 3, Docket Nos. 50-269, 50-270, and 50-287, October 23, 1978.

\.)

15. TMI-1 Fuel Densification Report, BAW-1389, Babcock & Wilcox, June 1973.

l 8-20 Am. 16  !

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(3) Draft Technical Specification 4.9.1.3 would require valve verification (correct position and locked, if appropriate) p for valves in the emergency feedwater system, every 31 days.

V In addition, locked valves could only be maintained in an un-locked condition under administrative control.

(4) Draft Technical Specification 4.9.1.4 would require a test, each 18 months, of automatic pump start logic and automatic valve lineup following an emergency feedwater actuation signal. In l addition, the operability of the manual control valve station would be verified.

l (5) Draft Technical Specification 4.9.1.5 would require testing of  :

the E W injection valves on a quarterly basis. ,

(6) Item 10F of NRC's October 26, 1979 letter to Mr. R. C. Arnold I requires a, ". .. Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow ,

test would be performed to verify the normal flow path from the primary EW system water source to the steam generators. ,

The flow test should be conducted with E W system valves in their normal alignment." This test is incorporated in draft Technical Specification 4.9.1.6 where the term " extended cold shutdown" is interpreted as "a cold shutdown of longer than 30 days' duration." ,

(7) Existing Technical Specification 4.1.2 (Table 4.1-2), would require a functional test of the Backup Instrument Air Supply O System (backup air supply for the emergency feedwater control valves), every refueling period.

(8) Existing Technical Specification 4.1.1 (Item 50 in Table i 4.1-1), as modified, would require a check each shift, i monthly testing, and calibration each refueling period, for the emergency feedwater flow instrumentation. The " check" I and " test" surveillances would not be required when TAVG is less than 200*F since the reactor would be shutdown and this l safety function not needed.

(9) Existing Technical Specification 4.5.1.1, as modified would incorporate the motor driven feedwater pumps into the list of equipment whose operation is verified during the testing of the emergency diesel generators. In this case, an addi-tional test signal is required to start the pumps since the pumps do not actually start on loss of AC power (the actual start signal is on loss of main feedwater or loss of reactor coolatt pumps.) l Conclusion In conclusion, with regard to the mcdifictions to the emergency l feedwater systems and associated Technical Specifications O l 11-15 Am. 16 l

t-The post-accident monitoring _ instruments to be installed at TMI-l l

are responsive to the recommendations of the NRC and the ACRS. L O Evaluation  !'

Section 2.1.2.1 of " Report in Response to NRC Staff Recommended Requirements for. Restart of Three Mile Island Nuclear Station  ;

Unit 1" describes the post-accident munitoring instrumentation  !

to be. installed at TMI-1. The post-accident instrumentation, in l conformance with Regulatory Guide 1.97, consists of the following.  ;

f (1) Containment Pressure - the range will be - 5 psig to three t times the containment design pressure; (2) Containment Water Level - a narrow Mnge monitor will measure

, containment sump level while the vi e range monitor will measure from the bottom of the containment to a 10 ft.

level; (3) Containment Hydrogen Indication - continuous reading of the  ;

concentration of hydrogen in the containment, from 0 to 10%, i will be available in the control room;  !

(4) High Range Containment Radiation Monitor - two monitors with  !

a range to 107 R/hr will be provided; (

r (5) High Range Effluent Monitors: i 4

O (a) Undiluted Containment Exhaust - 105 pCi/cc l 4

(b) Main Steam Lines - 102 pCi/cc l I

(c) Auxiliary and Fuel Handling Building Exhaust 103 uci/cc

(

(d) Condenser Off Gas - 105 pCi/cc (e) High Range Effluent Radio Iodine & Particulate Sampling and Analysis - silver zeolite cartridges.

Although the above instrumentation does not actuate safety equipment, nor is it required by safety analyses, it is appropriate to provide l Surveillance Requirements to assure reliable post-accident performance {

of the instrumentation. Surveillance Requirements for post-accident i monitoring instrumentation is incorporated into TMI-1 Draft Technical l Specification 4.1.1 (Table 4.1-1) as follows:

q (1) Item 13 of Table 4.1-1, "High Reactor Building Pressure," is provided with a footnote to include the post-accident instru- l mentation in the existing containment pressure instrumentation '

surveillance program; (2) Item 28 of Table 4.1-1, " Radiation Monitoring Systems," is i provided with a footnote to include the post-accident instru- {

mentation, described in item (5)(a) thru (5)(d) above, in the i existing radiation monitor system surveillance program; i i

11-17 Am. 16

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By increasing the Low Reactor Coolant System Pressure setpoint l O from 1800 psig to 1900 psig, the reactor would trip earlier in the LOCA scenario and thus the decay heat would be slightly  ;

less when the ECCS functions. Increasing the Low Reactor Coolant System trip setpoint also has the effect of increasing the margin ,

to DNB following a trip on low pressure; the reactor would trip earlier on low pressure and thus the final minimum DNB would be i higher (more conservative) than if the reactor tripped at 1800 t Psig. l Conclusion With regard to increasing the Low 1.eactor Coolant System trip setpoint from 1800 psig to 1900 psig:

(1) The probability or consequences of accidents previously con-sidered have not increased. For any accident that involves a l pressure decrease, the reactor will trip earlier in the trans-  !

ient and thus the result of the accident will be more conser-vative.

(2) No accident of a type not previously evaluated, will occur. l The increasing of the Low Reactor Coolant System trip setpoint will not have any effect other than tripping the reactor at an earlier time in pressure reduction transients.

(3) No safety margins have been decreased.

O It is expected that for pressure reduction transients, DNB following the reactor  ;

trip will be higher (more conservativa) and for the LOCA, the ,

peak clad temperature and other system parameters will be  !

more favorable. I Based upon the above, we conclude that increasing the Low Reactor f Coolant System trip setpoint from 1800 psig to 1900 psig does not t involve unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). <

11.2.13 Post-Accident Pressure Temperature Limits ,

Introduction i

Item 2 of the IE Bulletin 79-05B addresses actions to be taken by i reactor operators following automatic actuation of the High Pressure Injection (HPI) system due to low reactor coolant system pressure. Based upon IE Bulletin 79-05B, continued operation of  :

HPI is based upon maintenance of a 50'F subcooling margin in the '

I reactor coolant system; however, 'Ihe degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be j limited by the pressure temperature considerations for vessel  !

integrity." The purpose of this section is to describe draf t  !

l Technical Specifications which incorporate pressure temperature  !

{} limits to be utilized following HPI initiation.

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11-23 Am. 16

Discussion During certain Loss of Coolant Accident (LOCA) conditions generated by small breaks in the Reactor Coolant System (RCS), the HPI system is relied upon to effect cooling of the core. Cold HPI water is injected into a relatively high pressure, stagnant, RCS loop which can potentially cause a reduction in the margin to brittle failure of the reactor pressure vessel.

The high differential temperature between the RCS and HPI water has two significant effects:

(1) As HPI water flows into the reactor vessel, it will cool the metal. This consequently reduces the fracture resistance of the metal.

(2) Thermal stresses will be developed because of rapid cooling of the inside metal surface (thermal shock). This stress i will be superimposed on the existing residual stresses and those stresses that are generated by internal pressure. This combined stress field is significant, especially at a tLae when the fracture toughness of the reactor pressure vessel has decreased.

Analyses were made (Reference 1,2) of the consequences of HPI ,

actuation on the TMI-1 reactor pressure vessel considering effects '

of:

(1) Neutron fluence on fracture toughness of the material. '

(2) Resulting stress field. '

(3) Linear elastic fracture mechanics techniques that are out-lined in Appendix A to ASME Code Section XI and Appendix i G to ASME Code,Section III.

The results of the analyses are incorporated in a pressure-tem a perature curve presented in Figure 3.1-la of draft Technical Specification 3.1.2.1, " Pressurization Heatup and Cooldown Limitation." Draft Technical Specification 3.1.2.1 requires Figure 3.1-la to be used following HPI initiation and until HPI is secured and the reactor is in cold shutdown. Existing pressure-temperature limits for heatup/cooldown and in-service leak and hydrostatic testing, which apply to normal operation and testing conditions, would not be effected by the additional limita tions. The additional pressure-temperature 13mitations only apply to transient and accident conditions which result  !

in HPI initiation.  ;

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11-24 Am. 16 l l

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Conclusion The use of the post accident pressure-temperature limits pre-sented in Figure 3.1-la of draf t Technical Specification 3.1.2.1 will assure that an appropriate margin to reactor vessel brittle f ailure will be maintained following accidents and transients resulting in HPI initiation.

[

References (1) B&W Evaluation of RV Brittle Failure Due to Injection of  !

Cold HPI Water During Small LOCA Events. June 13, 1979.

(2) New B&W study to be finalized 1st week of March,1980.

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11-25 Am.16 l t

Draft Technical i

Specification Corresponding [

i to Section 11.2.7 I

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3.4 DECAY HEAT REMOVAL - TURBINE CYCLE <

Applicability r'T V Applies to the operating status of equipment that functions to remove decay heat, utilizing the secondary side of the steam generators.

Objective To define the conditions necessary to assure immediate availability of  !

the auxiliary feedwater system and main steam safety valves.

Specification  ;

3.4.1 With the reactor coolant system temperature greater than 2500F, three independent steam generator emergency feedwater pumps i and associated flow paths

• shall bc OPERABLE with:

a. Two emergency feedwater pumps, each capable of being powered {

from an OPERABLE emergency bus, and '

b. One emergency feedwater pump capable of being powered from an f OPERABLE steam supply system. With one emergency feedwater pump or flow path

• inoperable, restore the inoperable pump or flow path to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With more than one emergency feedwater pump l or flow path

• inoperable, restore the inoperable emergency feedwater ,

pumps or flow paths

• to operable status or be suberitical within 1  !

(') hour, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i L

t Four of six turbine bypass valves are OPERABLE.

c.

j 3.4.2 The condensate storage tanks (CSTS) shall be OPERABLE with a mini- i i

mum of 150,000 gallons of condensate available in each CST. With  !

a CST inoperable, restore the CST to operability within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or  ;

be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, at least HOT l

, SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  !

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l "For the purpose of this requirement, an OPERABLE flow path shall mean  ;

an unobstructed path from the water source to the pump and from the pump '

to a steam generator. i 3-25 I

3.4.3 With the racctor coolant systcm temparcture gretter then 2500F, all eighteen (18) main steam safety valves shall be operable or, if any are not operable, the maximum overpower trip setpoint (see

() Table 2.3-1) shall be reset as follows:

Maximum Number of Maximum Overpower [

Safety Valves Disabled on Trip Setpoint l Any Steam Generator __ (% of Rated Power) 1 92.4 2 79.4 3 66.3 With more than 3 main steam safety valves inoperable, restore at least fifteen (15) main steam safety valves to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Bases f A reactor shutdown following power operation requires removal of core decay heat. Normal decay heat removal is by the steam generators with the steam  !

dump to the condenser when system temperature is above 250*F and by the decay ,

heat removal system below 250*F. Core decay heat can be continuously l' dissipated up to 15 percent of full power via the steam bypass to thee condenser as feedwater in the steam generator is converted to steam by heat absorption. ,

Normally, the capability to return feedwater flow to the steam generators is  ;

provided by the main feedwater system. (

C:) The main steam safety valves will be able to relieve to atmosphere the t

total steam flow if necessary. If main steam safety valves are inoperable, the power level must be reduced, as stated in Technical Specification 3.4.3, such that the remaining safety valves can accommodate the decay heat.

In the unlikely event of complete loss of off-site electrical power to the station, decay heat removal is by either the steam-driven emergency l feedwater pump, or two half-sized motor-driven pumps. Steam discharge  !

is to the atmosphere via the main steam safety valves and controlled at- ,

mospheric relief valves, and in the case of the turbine driven pump, '

from the turbine exhaust.(1) ,

f Both motor-driven pumps are required initially to remove decay heat  ;

with one eventually sufficing. The minimum amount of water in the  !

condensate storage tanks, contained in Technical Specification 3.4.2,  ;

will allow cooldown to 250*F with steam being discharged to the ,

atmosphere. After cooling to 250*F, the decay heat removal system is used to achieve further cooling.  !

An unlimited emergency feedwater supply is available from the river via i either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period of time.  ;

O 3-26

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k The requirements of Technical Specification 3.4.1 assure that before the reactor is heated to above 250*F, adequate auxiliary feedwater capacity g is available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven pumps (460 gpm, each) are specified. However, only one half-capacity motor-driven pump is necessary to supply auxiliary feedwater flow to the steam generators in the onset of a small break loss-of-coolant accident (Reference 2).

The requirements of Technical Specification 3.4.1 assure that at least 920 gpm is available at all times to both steam generators giving re-dundant capacity except for a limited time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to allow for com-ponent maintenance. Further degradation of the emergency feedwater system requires the reactor to be suberitical within I hour.

The feedwater line break accident performed for TMI-2 (Reference 3) shows .

satisfaction of core thermal power limits and reactor coolant system pressure limits assuming full auxiliary feedwater flow within 40 seconds. .

The Technical Specification 3.4.1 provides assurance that this flow will be available with automatic initiation following loss of both main feedwater pumps. [

REFERENCES (1) FSAR, Section 10.2.1.3 (2) " Evaluation of Transient Behavior and Small Reactor Coolant System -

Breaks in the 177 Fuel Assembly Plant," Volume I and II, Babcock and  !

Wilcox, May 7, 1979.

O (3) Three Mile Island Nuclear Station - Unit 2, Final Safety Analysis 1

Report, USNRC Docket No. 50-320. 1 f

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TABLE 4.1-1 (Contirned)

CHANNEL DESCRIPTION O CHECK TEST U

CALIBRATE REMARKS O

Steam Generator Water Level

~

38. W NA R

39. Turbine Overspeed Trip NA R NA ,

40. Sodium Thiosulfate Tank Level NA NA R Indicator

41. Sodium Hydroxide Tank Level NA NA R Indicator

42. Diesel Generator Protective NA N R R21aying l 43. 4 KV ES Bus Undervoltage Relays NA M(1) R (1) Relay operation will be checked (Diesel Start) by local test pushbuttons.

, 44. Reactor Coolant Pressure S(1) M R (1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or ,

Taves is greater than 200*F.

45. Loss of Feedwater Trip S(1) M(1) R (1) When reactor > 10% power.

46 Turbine Trip / Reactor Trip S(1) M(1) R (1) When reactor > 20% power.

47. Pressurizer Code Safety Valve and Electromatic Relief Valve delta P/ flow S(l) R R (1) When T. AVG is greater than 2000F.

48. Pressurizer Electromatic Relief Valve - acoustic / flow S(1) R R (1) When TAVG is greater than 2000F.

49. Saturation Margin Meter S(1) M(1) R (1) When TAVG is greater than 2000F.

50. Emergency Feedwater Flow Instru- NA M(1) R (1) Emergency Feedwater is not normally l' mentation in operation.

iS - Ezch Shift T/W - Twice per week R - Each Refueling Period jD - Daily B/M - Every 2 months NA - Not Applicable ,

!W - Wackly Q - Quarterly B/W - Every two weeks Ji - Monthly P - Prior to each startup

TABLE 4.1-2

() MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control Rods Rod drop times of all Each refueling shutdown full length rods [

2. Control Rod Movement of aach rod Every two weeks, when reactor Movement is critical i

3. Pressurizer Safety Setpoint* 50% each refueling period Valves

4. Main Steam Safety Setpoint 25% each refueling period Valves

5. Refueling System Functional Start of each refueling period Interlocks

6. Main Steam (See Section 4.8)

Isolation Valves

7. Reactor Coolant Evaluate Daily, when reactor coolant System Leakage system temperature is greater than 525'F

8. Charcoal and high DOP test on HEPA Each refueling period and at efficiency filters filters, freon test any time work on filters for Control Room, on charcoal filter could alter their integrity and RB Purge units Filters P

9. Spent Fuel Cooling Functional Each refueling period prior to ;

System fuel handling

10. Intake Pump House (a) Silt Accumulation- Each refueling period ,

Floor Visual inspection of (Elevation 262 Ft. Intake Pump House Floor 6 in.) (b) Silt Accumulation Quarterly

, Measurement of Pump House Flow

11. Pressurizer elec- Setpoint Each refueling period tromatic relief valve ,

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12. Back-up instrument Functional Each refueling period air supply system i

• The setupoint of the pressurizer code safety valves shall be in accordance

()

l with ASME Boiler and Pressurizer Vessel Code,Section III, Article 9, j l Winter, 1968. i j 4-8 l i

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4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE Applicability Applies to periodic testing requirements for safety actuation systems.

Objective i To verify that the Emergency loading sequence and automatic power trans-fer is operable. '

Specifications .

4.5.1.1 Sequence and Power Transfer Test L

a. During each refueling interval, a test shall be conducted to demon- I strate that the emergency loading sequence and power transfer is operable.

b. The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred ,

i to the emergency power as evidenced by the control board component operating lights, and either the station computer or pressure /

flow indication.

- M. U. Pump

- D. H. Pump and D. H. Injection Valves and D. H. Supply Valves i

- R. B. Cooling Pump I

- R. B. Ventilators l

- D. H. Closed Cycle Cooling Pump  !

- N. S. Closed Cycle Cooling Pump [

- D. H. River Cooling Pump  :

- N. S. River Cooling Pump [

- D. H. and N. S. Pump Area Cooling Fan  :

- Screen House Area Cooling Fan  !

- Spray Pump. (Initiated in coincidence with a 2 out of 3 '

R. B. 30 psi Pressure Test Signal.)

- Motor Driven Emergency Feedwater Pump.

4.5.1.2 Sequence Test j

a. At intervals not to exceed 3 months, a test shall be conducted i to demonstrate that the emergency loading sequence is operable,  ;

this test shall be performed on either preferred power or emer- )

gency power.

b. The test will be considered satisfactory if the pumps and fans listed in 4.5.lb have been successfully started and the valves i listed in 4.5.1.lb have completed their travel as evidenced O 6 7 the ce=eret 6 ra ce 9e == ti== tisat - a etts r the atation computer or pressure / flow indication.

i 4-39

Bases

() The Emergency loading sequence and automatic power transfer controls the operation of the pumps associated with the emergency core cooling system and Reactor Building cooling system. A successful test of the emergency loading sequence i and automatic power transfer is a prerequisite to any system test of the emergency core cooling system or reactor building cooling system.

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i References t (1) FSAR Section 7 I i

(2) FSAR Section 1.4 i

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4-40

4.9 EMERGENCY FEEDWATER SYSTEM PERIODIC TESTING Applicability ,

() Applies to the perodic testing of the turbine driven and two motor-driven emergency feedwater pumps, assocaited actuation signals, and valves. i Objective i

To verify that the auxiliary feedwater system is capable of performing its design function.

Specification I r

4.9.1 TEST  !

i 4.9.1.1 Whenever the Reactor Coolant System temperature is greater than 2500F, the emergency feedwater pumps shall be tested in the  ;

recirculation mode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3000. The test frequency shall be at least every 31 days +7 days of plant operction at Reactor Coolant Temperature above 2500F. '

4.9.1.2 During testing of the emergency feedwater system when the reactor is in STARTUP or POWER OPERATION, if one steam gen-erator flow path is made inoperable, a dedicated qualified in-dividual who is in communication with the control room shall be continuously stationed at the EFW manual valves. On in-struction from the control room, the individual shall realign O the valves from the test mode to their operational alignment.

4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be verified to be in the status specified in Table 4.9-1. ,

4.9.1.4 At least once per 18 months, during shutdown, verify that:

(a) each emergency feedwater pump starting logic actuates upon receipt of an auxiliary feedwater test signal, and (b) valves in the emergency feedwater flow paths

• actuate to their correct position on an emergency feedwater test signal and that the manual control valve statirn functions properly.  ;

4.9.1.5 On a quarterly basis, the valves which are a part of the emer-gency feed system discharge (EFV-30A and 30B) will be checked for proper operation by cycling the valve over its full stroke.

4.9.1.6 Prior to start-up, following a cold shutdown of longer than 30 days' duration, conduct a test to demonstrate that the motor driven emergency feed pumps can pump water from the CST to the steam generators.

• For the purpose of this requirement, an OPERABLE flow shall mean

() an unobstructed path from the water source to the pump and from the pump to a steam generator.

4-52

4.9.2 ACCEPTANCE CRITERIA

() These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components s have operated properly. j l

Bases f i

The 31 day testing frequency will be sufficient to verify that the turbine  !

driven and two motor-driven emergency feedwater pumps are operable and f that the associated valves are in the correct alignment. ASME Section XI Article IWP-3000 specifies requirements and acceptance standards for the testing of nuclear safety related pumps. Compliance with the normal acceptance criteria of IWP-3000 assures that the emergency feedwater  :

pumps are operating as expected. The test frequency of 31 days (nominal) -i has been demonstrated by the B&W Emergency Feedwater Reliability Study  ;

to assure an appropriate level of reliability. If testing under Article l IWP-3000 indicates that the flow and/or pump head for a particular j pump is not within the normal acceptance standard, Article IWP-3000 [

requires that an evaluation of the pump performance shall be completed  !

within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or the pump declared inoperable. For the case of the  !

emergency feedwater system, the system shall be considered operable if under the worst case single pump failure, a minimum of 500 gpm of emergency feedwater can be delivered when steam generator pressure is 1050 psig and one steam generation is isolated. A flow of 500 gpm, at 1050 psig ,

head, ensures that sufficient emergency feedwater, demonstrated to be [

acceptable for plant cooling requireme.nts under transient and accident O conditions, can be delivered to either steam generator flow path. The 18 f

[

month surveillance requirements ensure that the overall emergency feedwater '

system functional capability is maintained comparable to the original  ?

design standards. I i

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Table 4.9-1 Status of EFW Valves l

Valve No. Status CO-V-10A Open i CO-V-10B Open l EF-V-1A Open EF-V-1B Cpen EF-V-2A 9 pen l

EF-V-2B Open  !

MSV-2A Open f MSV-2B Open {

i EF-V4 Locked Closed *  ;

i EF-V5 Locked Closed

• l t

EF-V6 Locked Open* l EF-V10A Locked Open*

EF-V10B Locked Open*

EF-V-16A Locked Open*

EF-V-16B Locked Open* I EF-V-20A Locked Open*

EF-V-20B Locked Open*

i EF-V-22 Locked Open* I l

• Locked Valves, if Maintained in a status other than indicated, l shall be under the administrative control of a dedicated qual- 1 ified, individual who is in communication with the control room. 1 The individual shall be continuously stationed at the EFW manual i valves and, on instruction from the control room, shall realign the valves from the test mode to their operational alignment.

1 i

i Draft Technical O Specifications Corresponding to Section 11.2.8 i

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O O O TABLE 4.1-1 (Continued) '

CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS .

10. Flux-Reactor Coolant Flow S M R Comparator

11. Reactor Coolant Pressure S M R Temperature Comparator

12. Pump Flux Comparator S M R

13. High Reactor Building S M R (1) Includes post-accident monitoring Pressure Channel (1) instrumentation (a) -5 psig to three-times design pressure

14. High Pressure Injection NA Q NA Logic Channel

15. High Pressure Injection Analog Channels

a. Reactor Coolant S(l) H R (1) When reactor coolant system is Pressure Channel pressurized above 300 psig or Tav. is greater than 200*F

16. Low Pressure Injection NA Q NA Logic Channel

17. Lou Pressure Injection Analog Channels

a. Reactor Coolant S(l) M R (1) When reactor coolant system is Pressute Channel pressurized above 300 psig or Taves is greater than 200*F

18. Reactor Building Emergency NA NA Q

Cooling and Isolation System Logic Channel I

1

. ._ . - ..-- ,.,_.-m.._. _ - . , . . . _ . . - , 7 . . - - -, . .r._,, . - .. , - - - - . -c.,_-_. ,-74 -.,, --,r -v , . - . - , . - - - _ . - - - - - . - - - . - - , , , , _ _ . . . , _ , - , - - - , . -

O O O.

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

28. Radiation Monitoring Systems (1) W(2) M Q(3) (1) Includes post accident monitoring ins t rumentation (a) High Range-Containment (b) Main Steam Lines (c) Auxiliary and Fuel Handling Building Exhaust (d) Condenser Off Cas Exhaust (2) Using the installed check source when background is less than twice the expected increase in cpm which would result from the check source alone.

Bac'kground readings greater than this value are sufficient in themselves to show that the monitor is function-ing.

7 (3) Except area gamma radiation monitors y RM-G6, RM-G7, and RM-C8 which are lo-cated in high radiation areas of the Reactor Building. These monitors will be calibrated quarterly or at the next scheduled reactor shutdown following the quarter in which calibration would normally be due, if a shutdown during the quarter does not occur.

29. Hikh and Low Pressure Injection NA NA R l Systems: Flow Channels i

l . . .

O O O l-TABLE 4.1-1 (Continued) ~

CIIANNEL DESCRIPTION CllECK TEST CALIBRATE REMARKS ,

30. Borated Water Storage Tank Level W NA R Indicator

31. Boric Acid Mix Tank

a. Level Channel NA NA R

b. Temperature Channel M NA R

32. Reclaimed Boric Acid Storage Tank

a. Level Channel NA NA R

{ b. Temperature Channel M NA R

33. Containment Temperature NA NA R

34. Incore Neutron Detectors M(1) NA NA (1) Check f'unctioning; including functioning of computer read-out or recorder readout when reactor power is greater than 15%

35. Emergency Plant Radiation M(1) NA R (1) BEttery check Instruments

36. Strong Motion Accelerometer Q(1) NA Q (1) Battery check

37. Reactor Building Sump and NA _ NA R (1) Includes post-accident moni-Containment Level (1) toring Instrumentation  ;

(a) Narrow Range (sump) l (b) Wide Range (Containment) l T

T l

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ - - - - - _ - - - - _ . _ . _ _ . _ _ . _ _ _ - _ - - _ ~ . - - . . . . _ _ . ~ . _ . _ . _ . - _ _ _ _ _ _ _ _ _ . . _ _ _ - _ _ _ _ _ _ . , , . _ _ . . . , - _ . . . _ . _ _ _ _ _ . _ . , . - . _ _ . , . _ _ . _ _

O O O -

TABLE 4.1-1 (Continued)

OIANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS

51. Reactor Coolant Pump Trip S(l) M(1) R (1) When TAVG is greater than 2000F.

Trip-

52. Reactor Building Hydrogen S(l) M(1) R (1) When TAVG is greater than 2000F.

Concentration I

D d

_____m _ _ _ _ _ _ _ - _ m -__ __m_ _ .. ,- -e - - . - - , - . _ ,. . - - - ._..,.- - - . , .. . - . - . - - . , . . _ - , , , -. , , ,._.,m._ ,. ,. _..-..,,m . , . , - _ . . . _ . , , _, . . - _ , - . . , , _

r 1 .

I i Draft Technical Specifications Corresponding to Section 11.2.13 [

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3.1.2 PRESSURIZATION EEATUP AND COOLDOWN LIMITATIONS Applicabilty i

Applies to pressurization, heatup and cooldown of the reactor  !

coolant system.

Objective  !

i To assure that temperature and pressure changes in the reactor l coolant system do not cause cyclic loads in excess of design for i reactor coolant system components. l Specification  !

l 3.1.2.1 For operations until five effective full power years, the reactor  ;

coolant pressure and the system heatup and cooldown rates (with l the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1, Figure 3.1-la, and Figure 3.1-2 and are as follows:  ;

Heatup/Cooldown f

i Allowable combinations of pressure and temperature shall be to the  ;

right of and below the limit line in Figure 3.1-1. Heatup and cool- l down rates shall not exceed those shown on Figure 3.1-1.

()

l Heatup/Cooldown Following HPI Initiation Following HPI initiation, and until HPI is no longer needed, and the reactor is in COLD SHUTDOWN, allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-la.

Inservice Leak and Hydrostatic Testing i Allowable combinations of pres are and temperature shall be to the '

right of and below the limit line in Figure 3.1-2. Heatup and cool-down rates shall not exceed those shown on Figure 3.1-2. ,

3.1.2.2 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 100*F. I 3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 100*F in i any one hour. The spray shall not be used if the temperature dif-ference between the pressurizer and the spray fluid is greater than 430*F.

3.1.2.4 Prior to exceeding five effective full power years of operation, f Figures 3.1-1 and 3.1-2 shall be updated for the next service period

() in accordance with 10 CFR 50, Appendix G, Section V.B. The highest 3-3 4

-e w - , v,

predicted adjusted ref erence temperature of all the beltline materials l

~g shall be used to determine the adjusted reference temperature at the

/ end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification i 3.1.2.5.

Based on the predicted RT NDT af ter five effective full power years >

of operation, the pressure-temperature limits of Figure 3.1-1 and 3.1-2 have been established in accordance with the requirements of 10 CFR 50,  ;

Appendix G. The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAW-10046. The pro-tection against nonductile f ailure is assumed by maintaining the coolant pressure below the upper limits of these pressure temperature limit ,

curves.  ;

i The pressure limit lines on Figures 3.1-1 and 3.1-2 have been established considering the following:

a. A 25 psi error in measured pressure.

b. a 12*F error in measured temperature.

c. System presure is measured in either loop,

d. Maximum differential pressure between the point of system pressure  ;

measurement and reactor vessel inlet for all operating pump com-binations.

Following initiation of HPI, additicnal stresses are introduced as a Y result of the high differential temperature between the reactor coolant system and HPI water. Figure 3.1-la (5,6) is for use following HPI initiation, and until HPI is secured and the reactor is in COLD SHUT-DOWN.

i The spray temperature difference restriction, based on a stress analysis .

of spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit. Temperature re-  :

quirements for the steam generator correspond with the measured NDTT for '

the shell. , ;

References  !

(1) FS AR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415  ;

(3) FSAR, Section 4.3.10.5 P

(4) BAW-1439, Analysis of Capsule TMI-IE From Metropolitan Edison Company, Three Mile Island Nuclear Station - Unit #1, Reactor Vessel Materials Surveillance Program. '

() (5) B&W Evaluation of RV Brittle Failure Due to Injection of Cold HPI Water During Small LOCA Events. June 13, 1979.

(6) New B&W Scudy to be finalized 1st week of March,1980. l l

3-4 l

_ _ . . . - . ,, , - - - - - - w 1- -

O O O TO BE SUPPLIED LATER FIGURE 3.1-la REACTOR COOLANT SYSTEM IIEAT-UP/COOLDOWN LIMITATIONS FOLLOWING llPI INITIATION (APPLICABLE TO 5 EFPY)

'""3" ' c "'"" ' " " ""'""'

C) SUPPLTIENT 1. PART 1  !

QUESTION NO. SYSTEM / TOPIC SUBJECT 1 EN Auto Initiation '

2 EFW Manual Initiation 3 EFW Diesel Generator Loading 4 EW Flow Indicators 5 , EN Startup and Testing of Initiating  !

Circuits t 6 EFW Flow Indicators 7 EFW l

Testing of Manual Control  !

8 EFW Flow Requirements  !

9 EW Failure Mode of Control Valves 10 EFW System Evaluation l 10a EN Level Indication for CST r

10b EFW Pump Endurance Test t

10c EFW Flow Requirements 10d EFW Testing Requirements 10e EFW Water Supply I 10f EN Testing Requirements 10g EFW Flow to Depressurized Steam Generator l t

10h EFW Pump Protection 101 EFW Independence from AC Power l

10j EFW Turbine Steam Line Break  !

10k EFW Auto Initiation p 11 Pressurizer Separation of Class I E Circuits Heaters  ;

12 Power Assign- Electrical Diagrams for New or ments Realigned Equipment '

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O ouest'o= =o- srstzs/ roe c susazcr 13 PORV and SV Flow Indicators t 14 Pressurizer Adequacy of 126 kw Heaters 15 PORV and SV Flow Indicators in Procedures 16 PORV and SV Test Program I 17 Inadequate Use of Existing Instrumentation  ;

Core Cooling  :

18 Procedures Use of Instrumentation Discussed i in Question 17 ,

t 19 In-Core Minimum Operable Thermocouples l i

20 i Subcooling Description r Meter 21 Containment Reactor Coolant Pump Supporting Isolation Auxiliaries O 22 i

t Containment Identification of Valves '

Isolation 23 Containment Electrical Drawings Isolation 24 Containment Reactor Coolant Pump Supporting Isolation Auxiliaries I 25 Containment Spurious Initiation Isolation 26 Containment Automatic Reset I Isolation 27 Containment Diverse Signals i Isolation P

28 Containment Pressure Setpoint Isolation 29 TMI Chronology Internal Review 30 Changes to PORV Internal Review O na nish rre Trip Setpoint nt- '

11 i.

QUESTION NO. SYSTEH/ TOPIC SUBJECT l 31 Reactor Trips Provision for Safety Grade Circuits 32 Emergency Submittal of Procedures for NRC l Planning Review '

33 Reactor Coolant Provision for Safety Grade Circuits L Pump Trip -

34 Reactor Trip / Supplying Data Turbine Trip i i

35 Small Break Analytic Assumptions  !

LOCA  !

36 Small Break HPI " Split"  ;

LOCA  !

36a HPI Core Makeup Capacity Following HPI I Line Break l 36b HPI Venturi Locations and Crossconnect i

() 36c HPI Testing of Design Chanses 37 HPI Installation of Venturis 38 Technical Submitted to NRC Specifications t

39 Inadequate Core Instrumentation, Procedures, and Cooling Training  !

t 40 Shift Supervisor Requirements 41 Shift Foreman Requirements

[

42 Shift Technical Requirements, Duties, and' Authority Engineer l 43 Shift and Relief Schedule for Submittal  !

Turnover l 44 Control Room )

Schedule for Submittal j Access, Onsite '

Technical Support Center, and Onsite Operational Sup-Q port Center 45 Small Break LOCA Procedures i

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E l QUESTION NO. SYSTEM / TOPIC SUBJECT f

46 TMI-2 Accident Exam for Operators 47 OARP Audit Criteria for Passing Grade Exam 48 Training Information on Design / Procedure  ;

Changes i 49 OARP Integration of Objectives into [

Training Program l

50 OARP Use of Outside Organizations ,

51 Natural Guidance to Operators Circulation  !

52 OARP Use of Plant Parameters 53 Safety Related Operational Review  !

Valve Positions 54 EFW Valve Check after Maintenance or Surveillance ,

55 Natural Filling of OTSG ,

Circulation 56 Engineered Terminating Operation  !

Safety Features  !

t 57 Notification Guidance to Operators  :

of NRC  !

58 EFW Proper Valve Positioning l 59 Containment Guidance on Resetting Isolation j 60 Shift and Re- Status of Safety Systems lief Turnover 61 TMI-2 Accident Lecture for Operators and Main- t tenance Persennel 62 Shift Staffing Commitment for Two Licensed  !

Operators  !

63 Automatic Switch Description of ECCS Design f s

Over to Recircu-lation iv

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O QUESTION NO. SYSTEM / TOPIC SUBJECT ,

64 Reactor Building Description of Modification Spray 65 Radiation Pro- Membership in PORC tection and ,

Chemistry Supervisor 66 Radiation Pro- Minimum Qualification i tection and l Chemistry .

Supervisor f

67 Radiation Pro- Detailed Description i taction Program  ;

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SUPPLEMENT 1, PART 1 QUESTION:

O 10. The Naa ser1eeins and Orders Task verce review of ePeratins reactors in light of the accident at Three Mile Island Unit 2 recently identified additional requirements for auxiliary feedwater systems. In addit.'.on to the requirements identified in this letter, other requirements which may be app 11 cable to the subject facility are expected to be generated by the D211etin and Orders Task Force. Such requirements are those resulting from our review of the  ;

loss-of-feedwater event and the small break loss-of-coolant accident. Our specific concerns include system reliability i (other than auxiliary feedwater system), analyses, guide-  :

lines and procedures for operators, and operator training.

The design and procedures of your facility should be evaluated against the following requirements to determine the degree of conformance. Provide the results of this evaluation and an associated schedule and commitment for implementation of required changes or actions described in Appendix A.

RESPONSE

The results of our evaluation of the requirements specified in questions 10a through 10k are fo11owing. The results of our evaluation of Appendix A to your October 26, 1979 letter is attached.

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SUBJECT CONTENTS OF NRC OUESTIONS O. . SUPPLEMENT 1, PART 2 QUESTION NO. SYSTEM / TOPIC SUBJECT 1 EFW Additional Information on Questions 1 and 10k (Supplement 1, Part 1) 2 EFW Additional Information on Question 9 (Supplement 1, Part 1) i 3 EFW Additional Information on Question 8 l (Supplement 1, Part 1) i 4 EFW Additional Information on Question 4  ;

(Supplement 1, Part 1) l 5 EFW Qualification for EIW Flow Measuring l Devices 6 EFW Control Room Annunciation of Auto-matic Starting 7 EFW Additional Information on Question O 10b (Supplement 1, Part.1) 8 EFW Additional Information on Question 10e (Supplement 1, Part 1) 9 EFW Additional Information on Question 10g (Supplement 1, Part 1) 10 EFW Loading of Motor Driven EFW Pumps 11 EFW Tech. Specs. on Flow Capacity 12 EFW Passive Failure of Discharge Line 13 EFW Additional Information on Question 10a (Supplement 1, Part 1) 14 EFW Additional Information on Question 10 (Supplement 1, Part 1) 15 Reactor Trips Evaluation of Parameters 16 Reactor Coolant Additional Information on Question Pump Trip 33 (Supplement 1, Part 1)

/

I 17 Reactor Trip Additional Information on Question 34 (Supplement 1, Part 1) l 1

() SUESTION NO. SYSTEM / TOPIC SUBJECT 18 Pressurizer Additional Information on Question Heaters 14 (Supplement 1, Part 1) '

19 PORV and SV Test Program 20 PORV and SV Additional Information on Question [

13 (Supplement 1, Part 1) 21 Safety Re- Additional Information on Question [

lated Valve 53 (Supplement 1, Part 1) [

Positions ,

22 Safety Re- Independent Verification lated Valve Positions 23 EFW Additional Information on Question 58 (Supplement 1, Part 1) 24 Containment Additional Information on Question Isolation 59 (Supplement 1, Part 1) 25 Shift Implementation of Shift Supervisor

() Supervisor Program 26 Shift Tech- Accident Assessment Function nical Engineer 27 Shift Tech- Training nical Engineer 28 Shift and Re- Schedule for Development lief Turnover 29 Shift and Re- Evaluation of Effectiveness lief Turnover 30 Pressurizer Design Details Heater Change-over from Control Room, PORV and Block Valve Power 31 EFW Isolseion of Manual Control from ICS 32 Containment Electrical Drawings Isolation O 33 EFW Circuit Interaction 11

I f

O QUESTION NO. SYSTEM / TOPIC SUBJECT 34 Power Assign- Additional Information on Question 12 >

ments (Supplement 1, Part 1) [

i 35 Reactor Trip Basis for Bypass of Turbine /Feedwater I Reactor Trips (

36 PORV and SV Backup Position Indication j 37 PORV and SV Electrical Drawings for Position l Indication  !

, 38 ICS Vulnerability to Loss of Power  ;

h 39 Accident Use of RETRAN I Analyses '

40 Airborne Definition of " Clear A*ea" l Activity [

l 41 Ingestion LPZ Definition 42 EPZ Location of Schools and Hospitals O 43 Hospitals Distance from Site to Hospital [

\

44 Emergency Relationship to Participating i Planning Authorities 45 Small Break Procedures l LOCA  :

46 Emergency General Emergencies, Precautionary l Planning Measures 47 Emergency Classification System Used by PENA .

Planning i 48 Emergency Early Warning Methods and Protective Planning Action within EPZ l

~

1 49 Emergency State and Local Emergency Plans l Planning Cross Referencing with TMI Emergency Plan 50 Emergency Offsite ESC, Manning by State and Planning Local Authorities O 51 Emer enc 7

. Planning 1nfermation en Re. 1ator7 ceide 1.101 111

QUESTION NO. SYSTEM / TOPIC SUBJECT 52 Fuel Handling Design Details Building En-vironmental Barrier 53 Solid Radwaste Design Details System

• 54 TMI-2 Decon- Independence from TMI-l Systems tamination 55 Natural System Response to Filling of Circulation the OTSC l 56 Station Location of Monitors  :

Discharge i 57 Plant Venti- Leak-tight Testing lation System 58 Met-Ed Duties of the Senior Vice President Organization 59 Met-Ed Duty Station for the Vice President Organization Nuclear Operations 60 TMI-l Site Interaction with Unit 2 Organization 61 TMI-l Site Rad. Chem and H.P. Functions Organization 62 TMI Station Clarification Support Organi- -

zation 63 TMI Site Provide Additional Details '

Organization ,

64 TMI Site Staffing Details '

Organization ,

65 TMI-l Site Clarification Organization 66 TMI-l Site Reporting Chains Organization '

O V 67 TMI-1 Site Qualifications for Radiological Organization Controls Manager iv j l

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( ) QUESTION NO. SYSTEM / TOPIC SUBJECT 68 Radiation Pro- Role of the Auxiliary Operator l tection Monitor l

69 Fire Protection Role of the Tech. Analyst  !

Organization l

70 TMI-1 Site Delegation of Authority During  !

Organization Contingencies ,

71 TMI-1 Site Personnel Resumes '

Organization 72 TMI Station Duties of the Director, Technical i Support Organi- Function l zation .

73 TMI Station Duties of the Director, TMI-2 Support Organi- Recovery  !

zation l

74 TMI Station Staffing Level l

Support Organi- '

zation 75 TMI Station Staffing Experience '

Support Organi-zation ,

76 TMI Station Resumes Support Organi-  !

zation 77 TMI Station Responsibilities Other Than TMI-l Support Organi- ,

zation '

t 78 TMI Station Site Assignment Support Organi-zation 79 Emergency Regular Assignment of Emergency Planning Organization Members  :

80 Emergency Extent of Mobilization of Planning Emergency Organization 81 Managee of Conflict between Sections 4 and 5 Support Services O ana to81stics 82 Emergency Staffing Level for Onsite Emergency Planning Organizations '

V

( QUESTION NO. SYSTEM / TOPIC SUBJECT 83 Emergency Qualifications for Onsite Emergency Planning Organizations 84

~

Emergency Interfacing Between Onsite Emergency Planning Organizations 85 Emergency Mobilization Times for Onsite Planning Emergency Organizations 86 Emergency Description of Duty Sections Planning 87 Emergency Training for Members of onsite Planning Emergency Organizations t

88 Emergency Detsils of Long-Term Recovery Planning Organizations '

89 Emergency Provisions for Keeping the Emer-Planning gency Plan Up-to-Date 90 Single Definition for " Safety Grade" Failure O Criteria 91 Hydrogen Request for Additional Information Recombiner 92 Natural Additional Information on Question Circulation 17 (Supplement 1, Part 1) 93 Emergency Additional Information on Question Planning

• 18 (Supplement 1, Part 1) 94 Subcooling Additional Information on Question Meter 20 (Supplement 1, Part 1) 95 Inadequate Additional Measures '

Core Cooling i

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l SUBJECT CONTENTS OF NRC QUESTIONS  !

O SUPPLEMENT 1, PART 3 ,

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QUESTION NO. SYSTEM / TOPIC SUBJECT I i

1 HPI Additional Information on Questions i 26 and 37 (Supplement 1, Part 1) '

2 HPI Additional Information on Question  ;

36a (Supplement 1, Part 1) 3 HPI Additional Information on Question ,

36c (Supplement 1, Part 1) l t

4 EFW Single Failure Criteria for EFW j Flow Control Valves l 5 Post-Accident Unacceptability of Letdown hbnitor Sampling 6 Range of Provide Additional Information I Radiation l Monitoru  :

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( 7 In-Plant Provide Additional Information f Iodine  !

Monitoring l l

8 Solid Rad- Storage Capacity f waste System  !

I 9 Solid Rad- Process Control Program  !

waste System  :

i 10 Solid Rad- System Capacity l waste System '

11 Reactor Coolant Design Requirements Pump Trip  ;

\

12 ICS Address Recommendations of BAW-1564 13 TMI-1/TMI-2 Systems Other Than Radwaste and Separation Effluent Monitors i 14 Radiation Recent Audit l Protection I Program 1 PORY Automatic PCRV Block Valve Closure l (Enclosure 2) I i

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_