Text

Auxiliary Feedwater Flow Rates Following Loss of Main Feedwater (177 Fuel Assembly Plants).
	ML19210E144
Person / Time
Site:	Three Mile Island
Issue date:	05/23/1979
From:	; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML19210E139	List: ML19210E138; ML19210E140; ML19210E144;
References
	86-1102587, 86-1102587-00, NUDOCS 7911300544
	Download: ML19210E144 (22)
	v • d • e

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t AUXILIARY FEEDWATER FLOW RATE FOLLOWING A LOSS OF MAIN FEEDWATER (177 FA PLANTS) 86-1102587-00 Document Identifier s

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/ 1.0 PURPOSE 56-1102587-00 Several studies have been performed (References 1 and 2) for the B&W 177 FA plants on a loss of main feedwater transient. The studies to date have looked at normal feedwater actuation times and delayed auxiliary feedwater actuation times. The purpose of this sensitivity study is to evaluate the response of 177 FA plants to minimum and maximum AW flow rates following a LOW event from 100% full power. This study then defines the minimum auxiliary feedwater flow rate that can be initia-ted to one or both OTSG's and return the plant to a stabilized cooling mode without filling the pressurizer, reaching saturation conditions in the primary side, or lif ting the PORV following a LOW event.

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-00 2.0 ANALYSIS ASSUMPTIONS A survey of B&W 177 Fuel Assembly plants shows that the auxiliary feedwater flow rates for the 2772 Mwt units can vary from 500 gpm (one AFW pump) to 1600 gpm (all AFW pumps). The smallest capacity at any of the B&W 177 FA plants is a one AFW pump flow of 370 gpm at a 2568 Mwt plant.

The computer code used for this analysis is the CADDS code (6ee Sec-tion 3.2.1.1 of Reference 1). The model and analysis assumptions of Reference 1, Sections 3.2.1.2, 4.2, 4.3, and 4.4 are used for this analysis. !

Volume III of Reference 1 verifies that this analysis is applicable to raised as well as lowered loop plants.

CADDS is a single loop model and therefore, the case where the minimum AFW flow is to one OTSG only is not explicitly studied. The ability of the code to model this situation is the subject of a separate report (Reference 2). The auxiliary feedwater flow rate in gpm is input as a heat demand.

The % heat demand (fraction of BTU /sec removed assuming 100% is 2772 Mut) is derived by considering the enthalpy change .in 80*F,1050 psig auxiliary feedwate- saturated steam at 1050 psig.

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j 3.0 ANALYSIS RESULTS 86-1102587-00 -

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/ A spectrum of auxiliary feedwater flow rate from 370 to 1500 sps were evaluated, for both trip on high RC pressure (2300 psig) and for the anticipatory trip at zero. The results, in terms of time to reach steam Fenerator level control, are given in Table 3-1. The cooling rate for each case is also presented.

In each case, the system parameters remained within normal ranges for a reactor trip. For example, the most severe initial undercooling case of

. 370 gpm AFW flow with a reactor trip on high RC pressure of 2300 psig (at 8 sec), resulted in a peak 4CS temperature following trip of 583.5'F. This case also sho~wed that AFW flow balanced the decay heat addition at about 8 minutes, with system temperatures dropping slowly (see cooling rate in

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Table 3-1) thereaf ter. Overcooling continues until 1000 psig is reached in the steam generator and OTSG 1evel control occurs at about 40 minutes (24uu

,T.1 sec). . For the maximum overcooling case (1500 gpm with anticipatory trip at

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'.'N. W.,. time zero), the steam generator pressure of 1000 psig is reached within about

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AUXILIARY FEEDWATER SENSITIVITY STUDY f

TIME OF REACTOR AFW FLOW TIME TO OTSG COOLING TRIP, SEC. RATE. CPM LEVEL CONTROL. SEC RATE, F/ MIN i .

0 370 1900 0.5 0 . 500 745 2 -

0 1000 215 . 3.5 0 1500 125 11 ..

48 (Reactor trips' 370 2400 0.5 :.

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48 Pressure) 1500 ' 150 11 5

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Aur.iliary feedwater flow rates of 370 to 1500 gpm provide satisfactory Performance in that RCS parameters remain within normal ranges. For under-cooling flow rates between 370 and 750 gpm, the RCS.will not fill nor lift ,

the PORV allowing heat production to decay to a level where level control can take place. For overcooling flow rates (>750 gpm), the pressurizer does not empty prior to the control point being reached.

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UV 11V4 dot-00 REFERENCES

2. . B&W Report to the '4RC, May 7,1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant".

2. B&W Report, " Anticipatory Trip Sensitivities", May 21, 1979.

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4-

QUESTION

4. Your response to Question 4 is not complete. Provide "as built" drawings and qualification documentation for the existing steam generator Javel instrumentation.

RESPONSE

The "as-built" drawings and qualification documentation forthe existing steam generator level instrumentation will be submitted at a later date under separate cover.

1438 127

QUESTION

5. Your response to Que.stion 6 is not complete. Provide qualification documentation for the ne'< " safety related" control on EFW flow measuring devices.

RESPONSE

The qualification documentation for the EFW flow measuring devices will be submitted to the NRC under separate cover.

t438'128

QUESTION'

6. Provide design drawings for the modifications which provide for control room annunciation of all automatic start conditions of the EFW system.

RESPONSE

The design drawings for the modifications which provide for control room annunication of all automatic start condition of the EFW system will be submitted under separate cover during the first week in January, 1980.

1438 129

QUESTION

7. Your response to Question 10b is not complete. Provide the test plan (procedure) for the proposed 72-hour endurance test on all EFW pumps, including your acceptance criteria.

RESPONSE

The test procedure, including acceptance criteria, for the 72-hour endurance test of the EFW pumps will be provided for your review and concurrence prior to conducting the test. The test will be completed prior to start-up and the results of the test made available to the on-site NRC inspector.

} hh

QUESTION

8. Your response to Question 10e i.= not complete. Provide the revised procedures as indicated for assuring an EFh' water supply.

RESPONSE

Procedure revisions detailing the method of assuring an EFW water supply are in the draft stage and have not received the approvals required by Administrative Procedure #1001 and are not presentiv available. These procedures will provide the methods for utilizing 'he following sources of EFW supply:

1.) Condensate storage tanks 2.) Demineralized water storage tank 3.) Reactor Building Emergency Cooling River Water system via EF-V4 and EF-V5.

These procedure revisions will be provided upon completion of the review and approval process.

e

QUESTION

9. Your response to Question 10g is not complete. Provide "as-built" drawings and a description in the Restart Report for the existing design for terminating EFW flow to a steam generator when low steam line pressure is revised.

RESPONSE

The steam line rupture detection is shown on elementary wiring diagram SS-209-143 and SS-209-144. Each steam generato* _s equipped with four (4) pressure switches, specifically PS 600, PS 601, PS 602 and PS 603 for steam Generator "A" and PS 604, PS 605, PS 606 and PS 607 for steam generator "B".

The steam line rupture detection system is designed to actuate when the steam pressure drops below approx. 600 psig. The actuation is on a steam genera-tor basis. The actuation system associated with one steam generator is independent of the other steam generator.

The following is a description of the actuation asscciated with steam generator "A". The actuation associated with the steam generator "B" is similar.

Referring to electrical elementary diagram SS 209-143, the steam line Al is monitored by PS 600 and the steam line A2 is monitored by PS 601. These two pressure switches control solenoid valves EF-V30A (SV1), FW-V17A (SV1),

FW-V16A (SV1), EF-V30A (SV3), FW-V17A (SV3) and FW-V16A (SV3). On low steam generator pressure these solenoid valves energize. These valves are piped in series in the " signal air line" between the E/P converters and valve positioner of the EF and FW valve.

Energizing both solenoids results in dumping the "s1 nal air" to the valve positioner to atmosphere, resulting in valve closure. This pneumatic anding of the solenoid provides for reliability and en-line testing one solenoid at a time.

To meet the single failure, a redundant set of pressure switches and solenoid valves are provided as shown on elementary wiring diagram SS 209-144.

Separate D.C. power supplies are provided. Separation of cable and seismic qualification are also provided.

The design meets the single failure requirement of IEEE 279-1971 and allows for on-line testing.

Bypass / defeat of the signal is required for normal shutdown. Switches 43A (#952) and 43A (#953) are provided. An alarm soundsif any defeat switch is actuated.

An alarm is also provided to indicate any pressure switch actuation. Indicating lights monitor each pressure switch circuit.

Limit switches on valve FW-V16A and FW-V16B are used in the control circuit of valve FW-V5A and FW-V5B.

1438 132

QUESTION

10. You have committed to modify the EFW system by providing automatic loading of the motor driven EFW pumps on their respective diesel generators during a loss of offsite power with coincident ESAS actuation condition. You have not provided sufficient justification to support this coincident logic. It is our position that adequate emergency feedwater flow for postulated accident conditions can best be assured if the motor driven EFW pumps are automatically loaded to the diesel generator on all loss of offsite power conditions.

RESPONSE

Emergency feedwater flow will be automatically loaded to the diesel generator on all loss of offsite power conditions. See Section 2.1.1.7.3, Amendment 6.

1438 133

QUESTION _

11. You have committed to provide a new Limiting Condition for Operation (LCO) in the plant technical specifications requiring an EFW flow path to each steam generator be available at 100% capacity. Your commitment states:

"If a flow path becomes unavailable or if capacity drops below 100S. to ,each steam generator, the plant shall be shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> ...."

To meet the intent of IE Bulletin 79-03A, Item B, it is our position that the word "each" be changed to the word "either" for this part of the LCO.

RESPONSE

See Section 11.2.1 as revised by Amendment No. 6.

)hb0

QUESTION

12. We have noted that the EFW pump discharge line cross tie contains two normally open motor operated valves. Upon loss of offsite power or loss of main feedwater, a single passive failure, such as pipe rupture in one EFW discharge line, could render both EFW trains inoperable. Provide the necessary modifications and/or procedural revisions to correct this condition and mitigate its potential adverse effects to plant safety.

RESPONSE

The subject piping is not high energy piping within the meaning of the Standard Review Plan Sections 3.6.1 and 3.6.2, therefore, rupture of the pipe is only possible during system use or during system surveillance.

Such a passive failure during system use (which is caly under emergency conditions) is not a design basis for the EFW system and is considered too improbable to warrant mitigation. EFW system surveillance is performed monthly and includes pressurizing this piping to maximum EFW pump discharge head. If a failure were to occur, it wctid therefore be discovered during testing and not under emergency conditions.

The EFW system is not used for plant startup.

(Also, see the response to Question 14 in Supplement 1, part 2).

~

1438 135

QUESTION

13. Your response to Question 10a provided in Amendment 5 is not complete.

You have not indicated whether the redundant condensate storage tanks level indication and alarms are powered from separate power supplies nor have you indicated that the low level set point provides at least 20 minu: s for operator action to assure an EFW water supply assuming that the largest capacity EFW pump is operating.

RESPONSE

At present, each condensate storage tank has a Low Level Alarm which is set to ensure that each tank level will be maintained above tech spec limits. A Low-Low Level Alarm will be added to each tank to give the operator at least 20 minutes warning before operator action is required to transfer to an alternate water supply. The indicators and alarms for the redundant tanks will be powered from separate power supplies. These modifications are planned as part of the long term upgrading of the Emergency Feedwater System. See response to Question 14.

1438 136-

QUESTION

14. Your response to question 10j provided in Amendment 5 is not complete.

Provide legible arrangement drawings for the EFW system showing the location of all system pumps, piping and valves. Provide qualifi-cation documentation which assures that the motor driven EFW pumps will start and remain operational under the environmental conditions (humidity and temperature) resulting from a postulated break in the main steam supply line to the turbine driven EFW pumps. Further, verify that the EFW control valves and actuators are qualified to function under these environmental conditions. Also, provide an analysis which justifies the environmental conditions (323*F) assumed as a result of the postulated steam line break.

RESPONSE

As described in response to Question 10j (Supplement 1, part 1) the subject break was not considered probable enough to warrant detailed design censideration at the time TMI-1 was licensed. Since TMI-1 was licensed, NRC acceptance criteria for EFW systems has been modified and the EFW system has taken on new importance. In recognition of this fact, Met-Ed has initiated a complete design review of the EFW system to upgrade it to the current licensing criteria to the extent practicable. This review will consider and resolve the type of concerns raised by questions 12,13 and 14 above. We believe that this approach is preferred over an item resolution of issues. Nevertheless, a response to your specific concerns is given below.

The EFW pumps have been certified to withstand the calculated environment.

A copy of the motor qualification certification is attached, together with the calculations which support the environmental conditions (323*F).

Environmental qualification of EF-V30A/B to 323*F was not invoked as part of the original purchase order for these valves, however, efforts are under way to determine if these valves can be certified to withstand the accident environment.

Arrangement drawings showing the location of important EFW valves and piping will be provided in about one week.

1438 137 1

, . 4 1

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RECElVED Mr5.1M3

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$ JUL0 1973 I

J M._ SAILER '

ANAT,YSIS OF INTER!EDIATE BUILDING EiVIRO:;! GIT

,, FOLLOWING A POSTULATED MAIN STEAM BREAK FOR THREE MILE ISLAND NUCLEAR PLANT 1

J I

]

Prepared for Gilbert Associates Incorporated 3

by

! Nuclear Services Corporation CAbfPBELL, CAUFORNIA i

e Prepared by: a . \.4 a- Issued by: 4,v>7 . --f4 DN R. A. Uffer ,( h R.'E. Kcever' Approved by: / ^c [ ' N Date: [ ' [ - [ 7 R. Bro 1438 138

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Nuclear Services Corporation TABLE OF CONTE'iTS

1.0 INTRODUCTION

1 Figure 1-1 2 Figure 1-2 3 h

2.0 ANALYTICAL DISCUSSION 2.1 System Description h 2.2 Analytical Methods 5 2.3 Assu=ptions ,

6 ,

Table 2-1 9 Figure 2-1 10

' Figure 2-2 11 3.0 RESULTS AND DISCUSSION 12 3.1 FRTHRUST Analysis 12 32 CONTD!PT Analysis 13 Figure 3-1 14 Figure 3-2 15 Figure 3-3 16 Figure 3-4 17 Figure 3-5 18 h.0 REFERENCES 19 i .

l .

i .

1438 139' .

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4 Nuclear Services Corporation a

1.0 INTRODUCTION

~

This report, prepared for the Gilbert Associates' Incorporated, presents the results of analyses predicting the environmental effects of a postulated main steam line break inside the intermediate building for the Three Mile o

Island Nuclear Station - Unit 1. The analyses included thermal hydraulic j blowdown analyses to calculate the mass flow rate and enthalpy of the i

fluid leaving the break, and time history building temperature and relative humidity calculations. The analyses cctnsidered only circumfer-ential breaks, since this type of break gives the maximum total flow and most severe conditions in the building.

The configuration of the piping and the postulated rupture locations are shown on Figure 1-1. The compartment configuration in the intermediate building is shown on Figure 1-2. A description of the analytical pro-a cedures and the analytical results, in terms of building'te=perature and relt.tive humidity vs. time after the break, are presented on the following pages.

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THREE ' MILE ISLAND INTERMEDIATE BUILDING .

SUBCOMPARTMENT LAYOUT \438 141 Nuclear Services Corporation iff

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I Nuclear Services Corporation 2.0 ANALYTICAL DISCUSSION 2.1 System Descriution ,

2.1.1 Main Steam System -

The main steam system consists of the piping used to deliver steam frw. the two steam generators to the main generator high-pressure turbine. It also .,

includes piping to provide steam during startup, shutdown, and normal operation to the main feedvater pump turbines, and during blackout or failure of both main feed pumps, to the emergency feedvater pump turbine which exhausts to atmosphere. There are two main steam lines frcm each steam generator to the main generator turbine or a total of four lines.

The only cross connection between lines is in the turbine steam chest between the turbine stop valves and control valves (=ain steam isolation valves). The turbine stop valves are both quick and tight closing; the control valves are quick closing. The piping arrangement is such that the rupture of a line from one steam generator vill not blev down the other steam generator,thus ensuring that a stea= supply is available to the emergency feedvater pu=p turbines.

2.1. 2 Main Steam Piping Analyzed The main steam r.iping considered in the rupture analysis consists of the -

sections of the four main steam lines (lines ME-38 -39, -40, -bl) between penetrations numbers h19, 114, 113, 112 respectively, and the main steam stop valves in the turbine building. All four lines enter the 3 _- --

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l intermediate building at the reactor building contain=ent vall. Inside the I

intercediate building the piping passes through several vall openings, then

{ penetrates the intermediate building vall to the turbine building through 36 inch diameter vall sleeves.

2.2 Analvtical Methods The blowdown analyses vere performed using the transient ther=al hydraulic' computer progra= PRTHRUST (Reference 1). This program models a physical system as an assemblage of volumes interconnected by flow paths.

Characteristics of a volume include state of the fluid and possible energy addition. Volu=es are used to model the steam generators, feedvater heaters and piping volumes. Flow paths are used to interconnect volumes, and can include operable valves, check valves, fills, and pu=ps. The program allows the operation of the valves, heat exchangers and fills to be triggered by time or by a physical signal suen as pressure. Subcooled break flow is calculated using the unsteady state Bernoulli equation and saturated break flov is calculated using Moody's critical flow model. Results of the analysis include ' ice history values of break flow, enthalpy and other thermodynamica1 quantities. .

The temperature and humidity of the intermediate building are evaluated by another computer program CONTDTT (Reference 3) . CONTEMPT predictr the pressure-temperature response of a building due to a fluid line rupture.

The input conditions for the program are the flow rate and the enthalpy of fhe leaking steam. The building volume is separated into a liquid region 1438 143

1 Nuclear Services Corporation i

and a vapor region. Each region is assumed to have a unifcirm temperature, but the te=peratures of the two regions may be different. The building is represented as consisting of several heat-conduct.ng structures whose thermal behavior can be described by the one-dimensional multi-region heat-conduction equation. The program also includes the building leakage through venting areas.

2.3 Assumntions 2 3.1 PRTHRUST analysis The main steam line rupture model for blowdown analysis using PRTHRUST is shown in Figure 2-1.

The conditions assumed for the analysis were:

a. The reactor is operating at full power until react'r o trip occurs, after which the reactor power is assumed to be given by the decay heat curve (Table 2-1).

b. A reactor trip occurs nine seconds after break.

c.

The turbine stop-valves close 1.1 seconds after reactor trip.

d. The steam generator pressure is 925 psia and the st'eam is saturated vapor. !

e. The feedvater pressure is 1050 psia and its temperature is h62 F

f. The break was assumed to be a guillotine severance of one steam line, such that flow issued from both sides of the broken line. .

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1438 144

j .

J Nuclear Services Corporation g.

Only the circumferential breaks were considered because this type of break results in maximum flow into the intermediate building.

h. Flov into the steam generator was based upon the feedvater pu=p performance curve.

i'. Failure of the feedvater control valve to close was assumed as the single active component failure.

J. Flav stops 600 seconds after the break when the condenser hotvell, which is the source of feedvater flow to the steam generator, is er.ptied. -

k. The effect of both steam generators and their associated lines is '

considered. ,

1. Both sensible and decay heat in the primary system are considered.

2.3.2 CONTDLDT Analysis The conditions assumed for the analysis were:

a. The intermediate building was codeled as one co=plete volume.

b. The valls and floors were modeled as exposed concrete surface to simulate their heat sink effect. The interior valls and floors were assumed heated on both sides while :xterior valls were heated on the interior side only.

c. Mass flow rate and enthalpy results from the PRTHRUST analysis were used for input to the CONTD!PT analysis.

d. A relative humidity of 60% exists in the intermediate building at the time of the break with an initial temperature of 90 F. and pressure 14.7 psia. ,

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. 1438 145 .

e

I Nuclear Services Corporation .

e. The building vent area is 167 square feet and has a discharge coefficient of 0.60.

f. Outside air conditions are 90 F.

g. The evaporation rr.te from the condensed liquid in the buildin;. to the vapor in the building at the end of steam generator blowdown is 28 lbs/second.

h. The heat rate between the atmosphere and the liquid in the building is 4.5 X 10 BTU / F HP..

.1438 146 O

9 e

6

Nuclear Services Corporation -

TABLE 2-1 REACTOR DECAY HEAT vs TIME Time after Trip Decay Heat (sec.) (H (t)/H,)

.0 1.0

.is y 1.0

.85 95 1.0 92 1.2 .80 1.4 .60 1 75 .30 2.0 .27 ,

k.0 .22 s 50 .205 6.0 .20 70 .195 8.0 .190 .

100.0 0.0 1438 147

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3.1 PRTHRUST Analysis Folleving a main steam line rupture, steam flows out of both ends of the broken line. The resulting depressurization of the affected steam gener-ator is assumed to cause a reactor trip nine seconds after the break occurs.

The turbine stop valves close 1.1 seconds after reactor trip, causing flov from the turbine side of the break to stop after the steam in the line be-tveen the break and stop valve has been exhausted, approximately 12 seconds after the break. Turbine side leak flow and enthalpy are shown on Figure 3-1.

Flow continues out of the steam generator side of the break, decaying with pressure, until the steam generator is exhausted. Normally, the feedvater control valve for the affected steam generator would close due to the steam line depressurization. The failure of the feedvater control valve to close is assumed to be the single active co=ponent failure for this accident.

When this failure occurs, the steam generator vill not exhaust as it is continually being fed by the feedvater system until the condenser hotwell is emptied 600 seconds after the break.

Initially, the fluid leaving the break is saturated steam. As th.e pressure

- decays, the enthalpy also lovers; the fluid, however, is still s',eam.

Following the break, the water level swells in the steam generator, causing water to be introduced into the steam lines. When this water reaches the 1438 150

Nuclear Services Corporation -

break, fluid enthalpy drops and the flowing 1 caving the break is two phase.

The fluid enthalpy then rises as the steam generator dries out and lovers again as the reactor heat decays. Steam generator side leak flow and enthalpy are shown in Figure 3-2.

3.2 CONTEFT Analysis Figures 3-3 and 3-4 show the te=perature versus time for the single volu=e building environment. A maximum temperature of 323 F occurs at one second, decaying to a 212 F pure steam condition after about one minute. After blevdown ends at ten minutes, air drawn in from the outside results in gradual cooling of the building. The atmosphere returns to 100 F after about h0 minutes, although the liquid inside the building is still at about 120 F. I'he relative humidity drops initially from the 605 value due to the temperature increase caused by high energy steam addition and super-heat buildup. After about one minute, the relative humidity ' reaches 100% due to the presence of saturated steam. Figure 3-5 shows the relative humidity transient following the accident.

e 1438 151 Nhclear Services Corporation l ply 1200- cau,nsix. cauronnis FIGURE 3-1. LONG TERM BLOWDOWil 1100- THREE MILE ISLAND MAIN STEAM LINE TURBINEGENERdTORSIDE (MASS FLOW RATE AND ENTHALPY vs. TIME) 1000 f ENTHALPY 900-800- 8000

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FIG.3-3. TEMPERATURE VS. TIME I

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Nuclear Services Corporation CAbtPDELL. CAUTORNIA FIG,3- 4, TEMPERATURE VS. TIME (LONG TERM) FOR TilREE MILE ISLAND INTERMEDIATE BUILDING 300- FOLLOWING. A STEAM LINE BREAK

-TURBINE SIDE BLOWDOWN 280- ENDS (10 SECONDS)

GENERATOR SIDE BLOWDOWN -

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h.0 REFEREIICES

1. Nuclear Services Corporation, "PRTHRUST: Cc=puter Code for Pipe Rupture Thrust Calculation," dated February 2, 1973.

2. Nuclear Services Corporation, " COMPRESSOR: Computer Code for Predicting the Temperature-Pressure Hirto:7 Within a Pressure-Suppression Contain-l ment Vessel in Responce to 2 Loss-of-Collant Accident.

i

3. Three Mile Island F.S.A.R.

i

4. Nuclear Services Report No. GIL-03-Oh dated March 12, 1973 - Evaluation of effects of Pipe Rupture Outside Containment for Three Mile Is1hnd

. Buclear Station - Unit 1.

i 1438 157 o

QUESTION

15. Provide your 2 valuation of anticipatory reactor trip parameters (feedwater pump turbine control oil rather than feed flow or other parameters). Include your evaluation of the need for a low steam generator level trip addressing various power levels. Discuss those transient scenarios that may not initiate anticipatory reactor trip for certain loss of feedwater/ condensate events (rather than high pressure reactor trip).

RESPONSE

A response to this question will be provided by November 30, 1979 or as soon as possible thereafter.

1438 158

QUESTION

16. Your response to Question 33 is not complete. Provide or reference a schedule for submitting the necessary analyses, operator guide-lines and revised emergency procedures required by Bulletin 05C.

In addition, demonstrate that installation of a safety grade automat ic pump trip is not practicable prior to restart.

RESPONSE

The Loss of Reactor Coolant Emergency includes instructions to trip all Reactor Coolant Pumps upon receipt of automatic actuation of High Pressure Injection (1600 psig). Guidance is being incorporated regarding restarting of the Reactor Coolant Pumps. These procedures are based on the Small Break Operating Guidelines dated November, 1979 and supporting analysis.

The analysis was submitted titled " Analysis Summary In Support of inadequate Core Cooling Guidelines for a Loss of RCS Inventory", document No. 86-1105508-

00. This document was sent to the NRC by Babcock and Wilcox and is intended to this requirement from NU REG 0578.

The modification to automatically trip the RCP's has only recently been initiated therefore the material requirements are not known. Without these details it is not possible to determine when the earliest possible date for installation might be. Preliminarily we believe that it can be accomplished by January 1981.

1438 159

QUESTION

17. Your response to Question 34 does not provide sufficient information for our evaluation. Provide:

A. Effects of liquid relief on the pressurizer electromatic relief valve following reactor trip no.11. Include results of any inspections conducted of the valve and discharge piping. Address valve performance and operability since this reactor trip.

B. Corrective action for main steam safety valves failure to rescat following reactor trip no. 12.

C. Corrective action for improper operator control of feedwater prior to reactor trip no. 11.

RESPONSE

A. Reactor Coolant leak rate records indicate no change in measured leakage after reactor trip no. 11. As a result it was concluded that no apparent damage occurred to this valve as a result of the trip. In addition main-tenance records on this valve indicate no repairs except for refueling interval routine maintenance have been performed on this valve.

B. Site records do not indicate that any corrective action for the main steam safety valves failure to reseat was taken as a result of reactor trip #12 nor has any further problem been experienced with the Unit #1 valves. In addition a review of the maintenance records indicate that no repairs of these valves have ever been performed.

C. No formal records of the action taken as a result of improper operator control of feedwater during reactor trip no.11 exist. However normal practice is to provde a critique of the reactor trip with each shift. It, therefore, can be assumed that each of the shifts were instructed on the mistakes made during reactor trip #11.

1438 160

QUESTION

18. Your response to Question 14 does not provide sufficient information for the staff to make an evaluation. Provide the calculations, assumptions and test data which justify the adequacy of 126 kw pressurizer heater capacity for maintaining natural circulation in the hot standby condition.

RESPONSE

The calculations, assumptions which justify the adequacy of 126 kw pressurizer heater capacity for maintaining natural circulation in the hot standby condition is stated in Section 2.1.1.3.1.2 of the Restart Report as modified by Amendment 3.

In addition, the data listed below is a summary of the heat loss through the insulation on the pressurizer which were taken during power escalation testing on May 20, 1974, at TMI. This indicates that the calculations were more conservative than the actual tests for heat loss through the insulation.

SUMMARY

- HEAT LOSS THROUGH INSULATION ON PRESSURIZER Average Heat Loss Sq. Ft. Ambient Surface Through Area Considered Area Temp. Temp. Insulation Area below Platform at El. 312'-7" 97 93* 105" 855.5 BT Area, El. 312'-7" to Bottom of Support Steel d El. 324'-0", Excluding Htr. Area 318 93* 110* 4460.0 BT Cylindrical Area (collar) Heater Bundle Insul. 43 93* 115' 823.0 BT Elev. Circular Area, Outside of Heaters 20 93* 140* 1010.5 BT Area, Hexagonal Part of Htr. Bundle Insul. 3 93* 200* 507.2 BT Area at Support Steel, El. 324'-0" to 327'-0" 90 110* 207.5* 12723.8 BT Area El. 327'-0" to 348'-0" 567 110* 152.5* 26748.2 BT Area, El. 348'-0" to Bottom of Top Head 108 110* 151.6* 4784.9 BT Vertical, Cylindrical Portion of Top Head 98 125* 152.4* 2550.9 BT Top Surface of Head Insulation 76 125* 220* 10721.7 BT TOTALS 1420 65185.7 BT Add Convection Losses 1893.2 2838.5 Total Heat Loss Through Insulation 69917.4 BT NOTE:

1. Operating temperature of pressurizer is 645'F.

2. Average heat loss through insulation: 69,917.4 = 49.2 BTU /Ft 2/Hr.

1420

3. Transco Reflective Insulation was 3" thick.

\458 \6\

QUESTION

19. As required by our October 30 letter, provide a description and schedule of the safety and relief valve test program. Verify that the program is applicable to U!I-1.

RESPONSE

As indicated in our response to Question 16 of Supplement 1, part 1, the subject relief and safety valve test program is to be conducted under the direction of EPRI/NSAC. The details of the test program have not yet been finalized. We expect that the details of the program will be available in early January 1980. We will describe the program details to the NRC at that time.

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QUESTION

20. Your response to Question 13 does not provide sufficient information for the staff to make an evaluation. Provide Appendix 2A to the Restart Report which contains calculations and test data for the PORV and safety valve flow indicators under various conditions.

With respect to the seismic and environmental qualification, the staff requirements are provided in the Lessons Learned clarification letter of October 30, 1979, pr.ges 7 and 8.

RESPONSE

The calculations and test data for the PORV and safety valve flow indicators under various conditions are not available. This information will be submitted at a later date under separate cover.

1438 163

QUESTION

21. Bulletin 79-05A, Item 5.

Your response to Question 53 addresses safety related valve position checks. IE Bulletin 79-05A, Item 5, also requires a review of the positioning requirements on all safety related valves. State your intent to compare the valve positions noted in your procedures (normal and maintenance lineups, checklists, etc.) with the process flow diagrams to ensure the procedures establish the proper flow path.

RESPONSE

Positioning requirements for safety-related valves will be compared with process flow diagrams to ensure procedures establish the proper flow path.

System Operating Procedures, surveillance test, and Special Operating pro-cedures for safety-related systems all require review by the Plant Opera-tions Review Committee and approval by the Unit Superintendent. One ele-ment of this review and approval process has been and continues to be the comparison of valve lineups with process flow diagrams to ensure applicable Technical Specification Limiting Conditions for Operation are observed.

The Engineered Safeguards checklist and auxiliary operator's logsheets' valve positioning requirements have been compared by the Supervisor of Operations Nith the PORC/ Unit Superintendent approved system operating pro-cedure valve lineups and with the system process flow diagrams to verify that the specified positions established the required flow paths, Post-maintenance value lineups of safety-related systems will be compared to system flow diagrams by two independent operator reviews of the switching orders prior to removal of safety tags to ensure proper restoration of the system to operability.

1438 M4

QUESTION

22. Bulletin 79-05A, Item 5.

Although IE Bulletin 70-05A, Item 5, does not specifically address an independent (second operator) valve alignment verification (reference Question 54), it is our position that such a check is necessary to ensure proper positioning after necessary manipulation.

State your intent to perform an independent valve alignment verifi-cation when returning the emergency feeC *ter system to operability after maintenance or surveillance testing. Also. submit your revised procedures reflecting this independent check.

RESPONSE

Met-Ed's acceptance of the NRC position on independent valve alignment verification is stated in memorandum GQL 1400, dated November 8, 1979 from J.G. Herbein to Richard Vollmer, NRC. This memo states, in part:

"At the completion of maintenance on ESAS and EFW Systems, an independent system checklist is required to return the system maintained to full operational status. This checklist will en-compass a specific system valve and breaker lineup within the boundaries in which maintenance was conducted. Additionally, this checklist will require two independent review and signature verifications for correctness prior to its use in aligning the system in the field. Following this dual verification, the checklist will then be used in the field to return the system to full operational status.

Following surveillance tests or special operations on ESAS and EFW Systems, two independent valve and breaker lineups will be conducted within the boundary of the system affected by the tests or special operations to provide assurance that the sys-tem is returned to full operational status."

The requirements for independent (second operator) valve alignment veri-fication for ESAS 6 EFW Systems are being incorporated into: 1) Admini-strative Procedure #1002, Rules for Protection of Men Working on Electri-cal & Mechanical Apparatus to require independent valve realignments upon completion of maintenance when safety tags are removed.

2) Administrative Procedure #1010, Technical Specification Survefilance Program to require independent valve realignments upon completion of surveillance tests.

3) Aiministrative Procedure #1001, Document Control, Appendix C, to re-quire independent valve realignments upon completion of special opera-tions on ESAS 6 EFW Systems governed by Special Operating Procedures.

1438 165

As these procedure revisions have not cocpleted the full process of review, approval and issue, they are not presently available for sub-mission to the NRC. They will be submitted as soon ss they have re-ceived the complete levels of approval required by Administrative Procedure #1001.

1438 166

QUESTION

23. Bulletin 79-05A, Item 7.

Your response to Question 58 (reference IE Bulletin 79-05A, Item 7) omits reference te operating procedure 1106-6, EFW System. State your intent to review this procedure and the others identified in your original response for proper positioning of the EFW system valves (i.e. , verify valve lineups).

_ RESPONSE Operating Procedure 1106-6 has been reviewed and procedural controls have been implemented to assure proper positioning of the EFW system valves.

Additionally, the other procedures identified in our original response to Questien 58 (Administrative Procedure 1012, Surveillance Procedures 1300-3F, 1300-3GAlB) have been reviewed and they also establish procedural controls necessary to assure proper valve positioning.

1438 167

QUESTION

24. Bulletin 79-05A, Item 9.

Your response to Questien 59 (reference IE Bulletin 79-05A, Item 0) did not include operator guidance for resetting of containment isolation caused by a . tor trip or a spurious actuation signal.

State your intent to provide this operator guidance.

RESPONSE

The design for containment isolation has been reviewed and it was determined th. 6 there is not automatic reopening of containment isolation valves upon reurtting of the isolation signal. Operator guidance for resetting the con-tainment isolation signal caused by reactor trip will be provided in the re-actor trip procedure (EP 1202-4).

Guidance for responding to spurious actuation of containment isolation will be contained in OP 1105-3 " Safeguards Actuation System".

1438 168 e

QUESTION

25. Your submittal on the Lessons Learned requirements on Shift Super-visor ResponsiSilities (NUREG-0578), Section 2.2.1.a, as clarified in our letter of October 30, 1979) is not complete since only Position 2 is addressed. Provide the details of the management directive (Position 1), training program (Position 3), and review policy (Position 4) used to implement this item.

RESPONSE

Position 1 Management Directive The management directive addressing the command responsibilities of Shift Supervisors in TMI-l is included as Enclosure 1.

Position 3 Training Program A five day decision analysis training program for Shift Supervisors which emphasized and reinforces their responsibility for safe operation and the Shif t Supervisors management function for assuring operational safety will be conducted. The training program objectives are:

A. To emphasize the Command role of Shift Supervisors in applying the Command responsibilities to operating problems.

B. To assure an understanding of how to apply basic decision analysis techniques to operating problems.

Additionally, the operator accelerated requalification program includes specific training which will also emphasize and reinforce the Shift Supervisor's responsibility for safe operation and the management function to assure safety.

Examples of this are:

A. Use of Procedures The shif t supervisors' role ir. the use of procedures is defined and discussed.

B. Procedures Review The shift supervisors, as well as the operators, are instructed in the content of emergency procedures.

C. Technical Specifications The responsibility of the shift supervisor end shift personnel with respect to Technical Specifications is stressed.

D. Safety Analysis Work Shop The team response to casualty situations is stressed.

1438 169

E. Simulator Training The team concept for casualty control was stressed. The shift supervisor was evaluated in his command role.

F. TMI Transient Constructive criticism of operator action during the transient was stressed in this portion of their training.

Position 4 Review Policy The administrative duties of the Shift Supervisor will be reviewed by appropriate members of the TMI Generation Group Staff in order to identify functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant.

The results of this review and recommendation will be documented by December 31. Appropriate recommendations will be approved by the Senior Vice President - Met-Ed, responsible for plant operations and will be implemented prior to 1 March, 1980.

1438 170 I

t I <

\l M

ENCLOSURE 1 TO QUESTION 25 p

T lij* f tszfj /((

kJ Metropolitan Edison Company Fast Office Box 480 Middletown, Pennsylvania 17057 717 944 4041 November 28, 1979 TO: SHIFT SUPERVISORS

TMI NUCLEAR GENERATING STATION

SUBJECT:

COMMAND RESPONSIBILITIES Nuclear generating facilities have the potential to signifi-cantly impact the health and safety of the public. This potential impact places a special burden and responsibility on those who manage and command operations at the Three Mile Island Nuclear Station.

The first line of defe.ase in portecting and assuring the health and safety of the public and the safety of personnel within the plant is the safe operation of all plant systems and components.

You, as the Shift Supervir' , have the primary management responsi-bility until properly relieved, for the safe operation of the plant under all conditions occurring on your shift. Accordingly, you are directly charged with both the responsibility and the command authority over all shift operations, radiological controls and maintenance activities under normal and abnormal conditions. Both the supervisor coming on shift and the supervisor being relieved shall make certain they review, convey and understand plant status and on-going activities and that the activities are deemed to be in c.ccordance with safety requirements.

Your responsibilities require you to constantly maintain the broadest perspective of operational conditions potentially affect-ing the general public, TMI personnel, and the safety of the plant.

Maintenance of this broad perspective shall be your highest priority at all times when you are on duty. In this regard, in times of emergency, you should be sure never to become so involved in any single operation that you are preoccupied to the extent that you might not provide adequate direction when multiple operations are required in the Control Room. During accident situations you shall remain in the Control Room to manage and direct the activities of the Shift Foreman, Control Room Operators, Shift Engineer, other plant operators and required support personnel until properly relieved.

\458 \1\

Metropohtan Edison Company is a Member of the General Pubhc Utaties System

l J N7 -f f jg Metropolitan Edison Company l g l [LJ I , e , ,). /

- h Jj Post Office Box 480 Middletown, Pennsylvania 17057

. 717 9444041 November 28, 1979 An essential element of protection of public health and safety is timely notification of State, NRC and Company officials in the event of an accident. There should be no reluctance on your part to initiate the notifications called for by the Emergency Pl?aa if conditions indicate a potential threat to public health or safety even if more evaluation is necessary to confirm the existence of such a threat. Further, it is imperative that you provide the opportunity for guidance and direction from the line management to which you report by prompt notification to them of the existence of abnormal conditions.

Constant, vigilant recognition of your management role to main-tain a command overview of the situation, to make decisions and to direct operations is the most important element in executing your responsibility to protect under all conditions, the health and safety of the public, the personnel on your shift, and the safe operation of plant systems and components under normal, off normal, and accident conditions.

J. G. Herbein R. /C. Arnold Vice President Senior Vice President

/

. y v -[f7/44- 7 H. Dieckamp President (Act g)

Ida 1438 172 Metropohtan Edison Company is a Member of the General Pubhc Utikties System

QUESTION

26. The Shift Technical Advisor discussion in NUREG-0578, Section 2.2.1.b (as clarified in our letters of September 13, 1979, and October 30, 1979) addressess accident assessment and operating experience assessment. Your submittal identifies the Shift Technical Engineer (STE) as providing the accident assessment function. State how you intend to accomplish the operating experience assessment function.

RESPONSE

Review of plant operating experience is a requirement that is placed on the Plant Operations Review Committee (PORC) by the TMI-1 Technical Specifications Section 6.5.1.6.f. To supplement the PORC review and to consider the experiences of other reactors a Plant Analysis Section has been established within the Technical Functions Division of the TMI Generation Group. This plant analysis section is responsible for reviewing all published LER's from other generating plants and determining the necessity for action at TMI-1. Although this group is located at the GPUSC home office, at least one of its members will be located at the TMI site to maintain close contact with the operations staff.

Information transmitted from the NRC through IE Bulletins, Notices, and Circulars relating to other reactors experiences is the responsibility of the TMI-1 on-site licensing group. This group is responsible for assuring that appropriate actions are completed on each IE transmittal.

They assign responsibility depending on the issues, and then track satisfactory completion of any necessary action items.

1438 173

QUESTION

27. Your submittal includes a proposed two year training program for the STE. This is not in agreement with the NUREG-0578 requirement for completion of STE training by January 1,1981. State your intention to complete STE training in accordance with the NUREG-0578 requirement.

RESPONSE

The Shift Technical Advisor training requirements as described on page A-51 of NUREG-0578 are:

"The shift technical advisor shall. . .have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout , including the capabilities of instrumentation and control in the control room."

The above training is included in parts I, II, III, and IV of the proposed training program and is scheduled for completion by January 1, 1981 in accordance with the requirements of Table B-1 in appendix B of NUREG 0578.

1438 174

QUESTION

28. Your submittal indicates that shift maintenance turnover procedures will be developed in accordance with NUREG-0578, Section 2.2.1.c, Position 2. Provide your schedule for development and implementation of these procedures.

RESPONSE

The procedure has been written and reviewed by PORC and implementation is expected by January 31, 1980.

1438 175

QUESTION

29. Position 3 of Section 2.2.1.c in NUREG-0578 requires that a system be established to evaluate the effectiveness of the shift turnover procedure. Describe the system you intend to use.

RESPONSE

The effectiveness of the shift turnover procedure will be evaluated by the use of the following:

1. Requiring applicable department heads to periodically review / sign their departments ' shif t turnover logsheets.

2. The Operation Quality Assurance Department will periodi-cally audit and review the effectiveness of the shift turnovers.

s 1438 176

QUESTION

30. The October 30 clarification letter concerning the Lessons Learned requirements identified two new items under Emergency Power Supplies.

The first item requires that the pressurizer heater be able to be switched to the emergency power supplies from the control room. The second item requires that the PORV and PORV Block Valve be pedered from separate divisions. Provide either modified designs or justifi-cation for the acceptability of your existing designs.

RESPONSE

The requirement to provide redundant emergency power for the minimun number of pressurizer heaters required to maintain natura: treulation conditions in the event of loss of offsite power does not co .and the degree of ready and available access provided by a control room located manual changeover of heaters from normal offsite power to emergency onsite power. At Three Mile Island Nuclear Station Unit 1, this changeover can take place up to two hours after loss of offsite power, thereby allowing ample time to perform the changeover remote from the control room. During the situation for which this changeover is required, there will be no restrictcJ access to the locations where the manual transfer is accomplished at the ES tusses.

The tsnual transfer scheme described in Section 2.1.1.3.1 provides positive physiaal circuit separation which prevents the inadvertant connection of one safety ius to the redundant safety bus or to the non-safety bus. A changeover of heaters manually performed in the control room that requires remote operation of non-safety equipment does not provide as high a degree of assurance against jeopardizing the safety busses.

The power to the PORV block valve is ABT protected and power is supplied from two vital busses. The PORV is also powered from the same vital bus that the block valve ABT is normally slected to, however, the PORV does not require power to close but rather closes upon loss or removal of power.

The objective of NUREG-0578 is, therefore fulfilled.

1438 177

QUESTION

31. Provide full details of the GE model CR-2940 switch to be used to isolate the EFW manual control station from the ICS.

RESPONSE

Full details of the GE Model CR-2940 switch to be used to isolate the EFW manual control station from the ICS will be submitted to the NRC under separate cover.

1438 178

QUESTION

32. Provide the electrical elementary drawings associated with the containment isolation modifications so that we may complete our review of this aspect of your design.

RESPONSE

The electrical elementary drawings associated with the containment isolation modifications will be submitted under separate cover during the first week in January, 1980.

1438 179

QUESTION

33. It appears from GAI Drawing SS-208-203 Rev. IA-1 that the manual and automatic initiation circuits are not arranged to prevent adverse interactions between them. If this is the case, provide justification for your design or modify it accordingly.

RESPONSE

The auto and manual initiating contracts are in p erallel since they both must actuate the "close" circuit of the EF" Pump Motor Circuit Breaker. Otherwise the functions are diverse and independent.

1438 180

QUESTION

34. Question 12 has not been fully answered. A listing of all new loads and bus assignments is required to provide an independent evaluation of whether the emergency power trains have maintained their independence.

RESPONSE

Some of the new loads have not been assigned yet. hhen these loads have been assigned, theinformation will be supplied to the NRC.

1438 181

QUESTION

35. Provide the basis for selecting 20% and 10% reactor power as the bypass setpoints for main turbine and main feedwater pump anticipatory reactor trip, respectively.

RESPONSE

A bypass of the turbine trip signal is automatically provided below 20%

reactor power in order to provide for normal starting and shut-down of the main turbine. There is adequate steam bypass capacity to accommodate a turbine trip at 20% power without enollencing the over-pressure trip function. A bypass of the FW Pump trip signal Les been provided in the design with a tentative set point at 10% reactor power. Further investiga-tion is being made to determine whether this latter bypass is required and what the set point should be. If it is determined that the bypass is not required it will be removed or the set point will be set to zero power.

1438 182

QUESTION

36. Provide a detailed description of the backup capability provided for determining the position of the PORV and pressurizer safety valves beyond the differential pressure transmitters.

RESPONSE

The differential pressure transmitters will be the primary indication that either the PORV or the safety valves are oper.. In addition, the PORV position will be monitored by the B&W supplied accelerometers which have previously been described in the Restart Report. Another diverse indication that either the PORV or the safety valves are open is the temperature detectors on the discharge lines of these valves.

We are currently completing analysis to document quantitatively how these detectors will respond at various valve openings and fluid conditions within the pressurizer. These calculations indicate that the temperature detectors will respond within a few seconds to the opening of a relief valve whereas response to the closing of a relief valve is several minutes. A plot of temperature detector response versus time for the various conditions analyzed will be prepared by January 15, 1980. This plot will be provided in the Restart Report and will slso be used to train the plant operators the proper inter-pretation of the readings from the temperature detectors. Opening of the PORV or pressurizer safety valves will result in changes in the Reactor Coolant Drain Tank. The tank level is recorded in the contrcl room and there is a high level alarm. There are also control room indicators for Reactor Coolant Drain Tank temperature and pressure.

1438 183

QUESTION

37. Provide detailed electrical drawings on the PORV and pressurizer safety valves position indication systems.

I'.ESPONSE The detailed electrical drawings for the PORV and pressurizer safety valve position indication systems were supplied to the NRC under separate cover.

The drawings were:

ECM 057 D-601-44-002 D-601-17-006 D-610-15-006 D-601-16-006 D-662-17-001 D-601-42-001 D-662-16-001 1438 184

QUESTION

38. With regard to a recent event at Oconee Unit 3 in which certain indications in the control room became unavailable, discuss the vulnerability of TMI-1 to a similar malfunction. Also, consider modifications which would reduce the potential for this type of event.

RESPONSE

The recent event at Oconee Unit 3 is being reviewed. If it is determined that 'DtI-1 could be subject to a similar malfunction we will proposed a suitable modification and the time frame in which it can be accomplished.

e 1438 183

QUESTION

39. Section 8 of the Restart Report indicates that additional transient and accident analyses will be performed using the RETRAN computer model. The connection of these RETRAN calculations to the previously reviewed BSh' calculations (Restart Report, Section 8, Reference 2) is not clear. Explain the connection and provide a list of all calculations currently planned using RETRAN. For each analysis, indicate the reason for the calculation. The concern is that RETRAN may be replacing previously accepted analyses, or may be used to support a design modification to a safety related system of interest to the staff. If so, the staff would require supporting documentation to justify use of this computer model. For example, benchmark comparisons to data and accepted models similar to RETRAN would be necessary.

RESPONSE

Section 8 of the Restart Report draws conclusions about the acceptability of the TMI-l design without making use of RETRAN analyses. The sources that were referenced to draw these conclusions include: the TMI-l and TMI-2 FSAR; Reference 2 of Section 8 of the Restart Report; and, the response to Question 3, Supplement 1, Part 2 of the Restart Report. Based on these sources, we are able to conclude none of the plant modifications to TMI-l result in a violatien of safety criteria; hence, they do not result in an unacceptable risk to the public.

RETRAN/GPU-01 has not been used to come to conclusions about the acceptability of the TMI-l plant design. However, the TMI-l computer model is being used to perform a number of accilent and transient analyses. These analysen will be used to expand the understanding of plant equipment performance during limiting and non-limiting accident and transient scenarios. RETRAN/GPU-01 has the capability of modeling plant control systems so that investigations can be made of systems such as emergency feedwater, turbine bypass and normal makeup during plant transients. The use of this detailed model provides plant specific informa-tion which is useful in operator training, evaluation of operating and emergency procedures, and design optimization. The use of RETRAN/GPU-01 does not replace the licensing basis analyses of the FSAR, or the abnormal Transient Operating Guidelines (ATOG) analyses being provided by B&W for the 177 FA Owner. Group.

RETRAN/GPU-01 analyses would not be used to justify design modifications which fell outside the limits of previously accepted analyses. If RETRAN/GPU-01 results contradicted accepted analyses, then these discrepancies would be discussed and resolved with B&W. In this respect, our analyses can be considered a tool used to review proposed procedures and designs.

In summary, the RETRAN/GPU-01 model will be used to: extend our understanding of plant response to various transients; aid in preparing operator training programs; and be used as a design and procedure review tool.

Appendix 8A to the Restart Peport provides some results of RETRAN/GPU-01 transient results. In the future, this Appendix will be expanded to include more modeling details as well as additional analyses. If practicab'.e, the Appendix will eventually include all of the accident / transient cases listed in Table 8A-1. l4jh jhh RETRAN/GPU-01 will not be used in support of licensing analyses used to support the initial restart of TMI-1. The model will be used in future licensing analyses.

References 24, 25 and 26 of Section 8 provide benchmarks of TMI-1 and TMI-2 plant tr msients to RETRAN/GPU-01 predictions. In addition, RETRAN/GPU-01 similations of the TMI-2 accident can be compared to other computer code predictions and to plant data. These results will be available in the near future.

QESTION

40. (1.7*)

Specify airborne activity levels for a clean area.

RESPONSE

A clean area is not defined in terms of airborne activity.

1438 18'7

QUESTION

41. (1.2.1)

Specify the ingestion EPZ.

RESPONSE

See Section 4.1.22, 1438 188

QUESTION

42. (2.1. 3)

Provide a figure showing locations of schools, hospitals, in the EPc_.

RESPONSE

See Tables 1, 2 6 3.

1438 189

QUESTION

43. (Table 3)

Provide distance from site to hospitals.

RESPONSE

See Table 3.

1438 190

QUESTION

44. (3)

Provide Tables similar to NUREG-0610 to indicate emergency classifications and their relationship to the participating authorities.

RESPONSE

See Tables 20, 21, 22 6 23.

1438 191

OUESTION

45. (4.1)

Define adverse meteorology.

RESPONSE

See Section 4.1.3 1438 192

QUESTION

46. (4.1)

Describe precautionary measures that may be taken in a General Emergency prior to a significant release.

RESPONSE

See Section 4.6.4.1 (2) 1438 193

QUESTION

47. (4.1 (3))

Describe the classification system used by PEMA and the counties.

RESPONSE

See Section 4.4.2 1438 194

QUESTION

48. (4.1(6))

Describe the methods of early warning of the public and the prompt initiation of protective actions within the EPZ.

RESPONSE

See Section 4.6.6 1458 140

QUESTION

49. (5.4)

Provide a specific cross reference to the information in State /

local plans requested in this section.

RESPONSE

See Section 4.5.2 and 4.5.3 1438 196

QUESTION

50. (5.3)

Describe the manning of the offsite EDC by Local and State authorities.

RESPONSE

See Section 4.7.2.1 1438 197

QUESTION

51. (5.3)

Provide the previously requested information for the following sections:

Response to R.G. 1.101 Regulatory Guide 1.101 See Emergency Plan Sections

5. 2 (1) 4.5.1.3 5.2.1(2) 4. 5.1. 3 (1) 5.3. l(3) 4.6.5.4

5. 3.1(4) 4.6.5.4 6.4 (1) 4.6.4; 4.6.6; 4.6.6.1; 4.6.6.2

6. 4 (2) 4.6.4.l(1); 4.6.6; 4.6.6.1; 4.6.6.2 6.4.1 4.6.4.l(1) 6.4.3.2 4.6.4.3 6.5.l(2) 4.6.5.1 6.5.4 (1) , (2) , (3) 4.6.5.4

7. 2 (1) 4.7.5 7.3.2 4.7.6.2 7.5(2) 4.6.5.4 8.1. 2 (7 ) 4.8.1.2 10(1)2 Figures 21, 22, and 23 Response to Review Guideline No. 1 Review Guideline No. 1 See Emergency Plan Sections IA6 4.7.2.1 IB4 4.7.2.1 IB5 4.5.3.4 IIAB 4.4.1 IIA 4 4. 7. 6. l (7 )

IIB 4 4.4.2 IIB 5 4.6.6.1 IIB 6 4.6.6.2 IIA 7 4.6.6.2 IIIAl 4. 7 . 6.1 (7 )

IVA1 Appendix C, Figures 20, 21, 22 and 23 VB2 4. 8.1. 2 (5)

VB3 4.5.4

)hb0

QUESTION

52. Provide detailed design features of Fuel Handling Building environmental barrier.

RESPONSE

The following describes the TMI 1 and 2 Fuel Handling and Auxiliary Building supply and exhaust systems. The potential leakage paths between buildings or systems and the modifications designed to isolate the unit I refueling floor from the unit 1 Auxiliary Building and from the Control Access Building are discussed. These modifications include ventilation system changes and certain building layout changes. The major ventilation considerations are as follows:

A. The supply and exhaust systems for unit 1 are separate from those of unit 2. However, the unit I refueling floor air communicates directly with the unit 2 refueling floor air.

B. The supply systems of the Auxiliary and Fuel Handling Building (FHB) of TMI-1 are separate from each other. Both systems supply air to the building areas through duct distribution systems using outside air drava from the air intake tunnel. Both supply fans are located in a common tunnel in close proximity to each other.

None of the supply ducts of the Auxiliary Building are located in the FHB area. Thus, there is no potential for air leakage between Auxili-ary and FH Building through e lets or through leaks in the Auxiliary Building supply duct system.

The supply duct main for the FHB serves the general area at elevation 305'-0", the Spent Fuel Cooling Pump area at elevation 305'-0" and then serves the refueling floor at elevation 348'-0". The FHB refueling floor could communicate with the Auxiliary Building through the supply duct system because the general area and the Spent Fuel Cooling Pump area are open to the Auxiliary Building through an open doorway at elevation 305'-0".

C. The exhaust systems for Auxiliary and FH buildings of TMI-1 are separate in the specific buildings they serve but the FHB exhaust main becomes common with the auxiliary building exhaust main after leaving the FHB.

The common main is directed to multiple filter plenums and fans that exhaust the mixed air.

The building modifications designed to isolate the TMI-1 refueling floor from the TMI-1 Auxiliary Building and from the Control Access Building include:

A. Two pairs of double doors at elevation 281'-0".

B. An enclosed passage at elevation 305 -0" with two main doors and one pair of equipment doors.

1438 199

C. A wall at the east end of the truck bay at elevation 305'-0".

The wall should be removable for large equipment access to the machine shop.

D. A security fence at the west end of the dock adjacent to the new enclosure at elevation 305'-0".

E. The stair tower between elevations 299'-2k" & 211'-0" will be modified.

F. Pressure tight doors for the new fuel storage room at elevation 329'-0".

G. Pressure tight doors for the stair tower at elevation 331'-0".

H. An enclosure at elevator entrance with a pressure sealed door.

I. An enclosure for the ventilation duct chase in the northwest corner of the refueling floor with one pair of pressure tight doors.

The TMI-1 ventilation system modifications designed to prevent the leakage paths are given below:

A. Air leakage from the FHB through the supply duct, to the de-energized FHB supply f an to the Auxiliary Building are stopped by adding a leak tight damper in the discharge of the FHB supply fan.

B. Air leakage from the FHB through the supply duct, and the 48 x 2k branch duct, to the FHB general area at elevation 305'-0", and then to the Auxiliary Building will be stopped by blanking of f this du:t and providing an equivalent opening in the FHB supply duct. This would discharge the required 8000 cfm on the south side of the elevator shaft and this air would rise through the open stairwell to be exhausted at the refueling floor.

C. Air leakage following the same path as item B above but through a 12 x 12 branch duct in the spent fuel pool cooler area at elevation 305'-0" will be stopped by blanking off this duct. The 1000 cfm exhaust required by this area will then be supplied from the Auxili-ary Building through the wall opening at elevation 305'-0".

D. Air leakage from the Auxiliary Building to the fuel building will be stopped by adding a leak tight damper in the exhaust duct main as it leaves the FHB but upstream of the connection with the Auxiliary Building main (60 x 50, elev. 348).

E. The leak tight dampers added to the FHB supply and exhaust ducts and the supply fan will function as follows:

1438 200

1. Automatically close on detection of differential pressure in the FHB with respect to the Auxiliary Building.

2. Automatically close on detection of high radiation on the refueling floor or in the FHB return air duct.

3. Open or close on manual command from control stations in the FHB and in the Control Room. The manual station would "over-ride" the above automatic signals.

The post modification system response following various assumed accidents and events is given below:

ACCIDENT / EVENT RESPONSE

1. Differential A differential between these two areas could develop pressure develops as a result of the loss of either the fuel building between the Auxili- or the auxiliary 'uilding supply fans or failure in ary and FHB. the closed position of the dampers in the exhaust mains from either of the buildings. Leakage between the buildings in either direction would be stopped by adding the building and the ventilation system modifications.

2. High radiation The ventilation system modification would automatically in the FHB or in isolate the TMI-l FHB from the Auxiliary Building.

the FHB exnaust Supply air flow and exhaust air dampers would initially duct. close and the FHB supply fan would stop. However, subsequently, these dampers could be opened and air flow through the fuel building could be re-established at the discretion of the plant operator. Also, the operator could limit the exhaust flow by opening only the exhaust damper. Operation of the Auxiliary Building supply and exhaust system would continue during the isolation phase.

3. High radiation High radiation would be detected and alarmed locally and in TMI-1 Auxiliary in the control room by area monitors and by a monitor in Building. the exhaust duct from this area. The supply and exhaust system could continue operation to reduce the radiation level. If necessary, the refueling floor could be manu-ally isolatad.

4. High radiation The same response as noted in item 3 above would occur.

in TMI-2 refueling floor .

[k)b .

In addition to the modifications described abave, a ventilation system to mitigate the consequences of a postulated fuel handling accident in the FHB will be installed unless it can be demonstrated that such a system is unnecessary. This new system will meet the requirements of Regulatory Guide 1.52, Revision 2.

Preliminary calculations performed in accordance with Regulatory Guide 1.25 indicate that of fsite doses will be a small fraction of 10CFR100 limits taking no credit for a ventilation system and assuming a ground level release.

Other assumptions used are that: (1) 56 fuel pins are assumed to fail, (2) the DF for water in the fuel pool is 100, (3) core power is 2535 Mwt, (4) a radial peaking factor of 1.65 is assemed, and (5) a site X/Q of 6.8 E-4 is used.

Final results of the f uel handling accident analysis will be provided by December 7,1979 and will form the basis for TM1-1 operation until the above mentioned ventilation system can be installed if it is necessary to install it.

1438 202

QUESTION S3. For Solid Radwaste Systems, provide the following:

A. Description B. Capacity C. Process Control Program D. On-Site Storage Facility E. Expected amounts of Solid Wastes Per Year.

RESPONSE

A. A backfit to the TMI-1 Radwaste Solidification System is planned and consists of replacing the existing Protective Packaging, Inc. (PPI)

System, which utilizes urea formaldehyde as the solidifying agent with one that uses a Dow Chemical Co. polymer as the binder.

The Dow waste solidification process will be used to solidify radwaste in the form of liquids, slurries, and spent ion-exchange resins.

Proprietary technology is used to form stable water-resin emulsions which are then chemically cured to form hard, solid, monoliths.

Liquid or slurry waste is stirred with a commercially available binder until a stable waste-binder emulsion is formed. The mixture is then cured by the addition of two chemicals. The final result is a dispersion of small spherical liquid particles (fine droplets) in a continuous matrix of cured binder.

An approved process control program will be incorporated into the system in accordance with Regulatory Guide 1.143.

The new system will be physically located in the 1111-1 Radwaste Solidification room where the PPI System is currently located.

Engineering for this backfit w111 begin in November 1979 and is scheduled for completion approximately April 1,1980. Equipment installation and checkout should be completed by October 1, 1980.

B. Capacity of radwaste solidifcation system:

200 cwft/ day . Solidi fied Waste 120 cwft/ day Raw Waste

( 900 gallons / day)

(Based on the operation of DON System).

C. The Process Control Program will comply with Regulatory Guide 1.143.

} k )b -

D. The solidified waste will be stored until shipment with the Epicore 11 wastes until a permanent waste storage building is available.

F- daticipat:2 amounts of solid radwaste produced per year:

5000 cwft Solidified Evap. Bottems 3000 cwft Compacted Trash (Dry) 1000 cwft Solidified Resin (Based on a normal operating year with refueling outage).

1438 204 4 m .-

QUESTION

54. Provide information confirming that the TMI-2 decontamination and restoration operations will neither depend upon the TMI-l radwaste systems nor will be interconnected with the TMI-1 radwaste system.

RESPONSE

Section 7.2 of the Restart Report provides information regarding the manner by which the radwaste systems for Units land 2 will be separated to prevent potential contamination of the unit. Upon completion of the currently planned activities, each unit will be completely separate and will be able to process all wastes associated with its own operation or recovery activities. All new processing equipment required to support the Unit 2 recovery operation will be located in Unit 2 facilities and all wastes generated will be handled in these facilities.

No additional connection to Unit 1 radwaste facilities will be made, and hence no potential for cross connection to Unit 1 will be available.

1438 205

QUESTION

55. Provide the TMI-1 sump pumping and sampling procedure (SOP Z-33) for industrial waste treatment facilities.

RESPONSE

See attached procedure.

1438 206

. Attachment to Question 55.

y. .

'#kP1001 Three Mile Island Nuclear Station $1DE 1 2

,.( . figure 10018 Special Operating Procedure [d'v S0P NO. 6 -33

, , p (From SOP Log inden)

NOT E: Instructions and guedelsnes in AP 1001 h musi u. ,ono..d is i.n compiet.na inis Unit No. Z~/ 22~

arm. raw w a#(,lai K ,l0 't*'lp Date 9/7/h 9

/

LA TE A

1. Title !"a pt p At Sc l 'A A rdC To .7~ta TS f 1'cJ M

2. Purpose (inciud. purpos. or SOP) g g , m 7!s **s 50f Ga pri < cons 2-t j..,qu i

- 1 i. < mov

.,s

3. Attach procedure to this form written according to the following intmat. - '

A. Limitations and Precautions " ' !

Nd 1.

2.

Nuclear Safety -

EnvironmentalSafety - ses < a.asa f

..(t- - ( i ( ,

,'r".

,,,, e.: J. a u a '. s r4id a bJ.;.g p --

.j of g g

3. Personnel Safety . gA

4. Equipment Protection - fa,e B. Prerequisites -f mrt c d er d i C. Procedure 5p E/9/f)

/ /t, / i

4. Generated by

[//v . b,[ -

Date N d7 h. ,h t[ 3.

%. ,' Duration of SOP . snaiso. noionsort o days vrom ine ef reetiv.dat. or tne sop or (a) or io> t.. sow - nichev.r occurs virSt.

(a) SOP will be cancelled by incorporation into existing or new permanent procedure submitted by C (b) S0 P is not valid af ter 9D /)4 v .5' &

(fill in circumstances which will result in SOP t>eing cancelled)

.G. .(a) is the procedure Nuclear Safety Related? '

If "yes", complete Nuclear Safety Evaluation. < car.2 oe inis eormi . .... . . . . Yes O No I (b) Does the procedure affect Environmental Protection?

If "yes", complete Environmental Evaluation. is.de 2 ot tn.s ro,,n> . . . . . . . . . Yes 3 No (c) Does the procedure affect radiation exposure to personnel? . . .

NOTE:

. . . . .. . . . Yes B' No if au answers are "no". tne cnenQe may be apgiroved by tne Shif t Supervssor. if any questions are answered "yes", tne cnsr se anust he approved by tne Unit Superintendent.

< jl

. 7. . Review and Apprnval / '

/ <

~

Approved - Shif t Supervisor ,

h([ [

Reviewed - List members of PORC contacted A. u,1res,f /,,./n f acus').' "* M. f jtfq fnc ,u ,u m s w t.i. f s o V '

o a t. '

'T. wi..a V /11 Sa uu %. /79

/Q C by TsI U *

h?

w. i n e n. w , % l:, Aert) o,,,%./79 k (* y

_,, c su ta va s w '

7,iu . n,6 %

<v~rm"^ r naaca n %/n A e col p ",* % . / ,9

! oe '

4 Approved - Unit erintendent Sup'4: I 4c m) fM/n I,/

(< 1[e: 3 y. ,, a y[/

C; Reviewe - Sup. Quality Controf tie r.auer.d> OfA Mb l oate

8. SOP is Cancelled Snitt Supervisor /Shif t F oremen o a t.

ld }Q[

                                                                                                                                                  .        1    n r *,       -

                                                                                  " EVALUATION"                                                                           -

Three flife Island Nuclear Station

             .' A P-1001
           ,, Figure 10018                                        Nuclear Safety /EnvironmentalImpact Evaluation SID E 2        /

SDP No. f- 33 '( \ -

1. Title

2. Nuclear Safety Evaluation Does this S0P

                            * (a) increase the probability of occurrence or t "e con %rij es of an accident or m'                          ction of equipment important to safety? . .                  .       .. . . . . . .                    . .     . . .         yes           no       .
                            * (b)     create     the    possibility         for   an   accident ctic  o'  a      orpalf different  typ un/.

than any evaluated previously in the safety analysis . . . . . . . . . repor 7 /'e/ .

                                                                                                                                   . . . . . . yes                     no O
                            * (c)     reduce the margin      of safety    as define    'in th ba     is f a ny    technicalspecification?        . . . . yes           no Details of Evaluation       tempi.in why .n. wor. io .bov. au.       ns a         *. Att.ch additionalbages if required.)
                                .                                                   J
              ~

Evaluation By Date

3. Environmental Impact Evaluation Does this SOP:

                            * (a) possibly involve a significant environmentalimpact? .                    . . . . . . . . . . ...                       yes           no (b) have a significant adverse ef fect on the environment? . . . . . . . . .                           . . .     . . yes               no O
                            * (c) involve a significant environmental matter or question not previously reviewed and evaluated by the N.R.C. . . . . . . . . . . . .. . . . . . . ...                                               yes           no T Details of Evaluation                                                                                 ,

7L's ja s an ou u w.?Il jaaneo Ao:toEsn( ,,, u , a ,a s tl.c 7 7 /, ..t J.sif .g > < mat duH -d, ,,pon

                                                                             ,,,,        yn             /. .~; U, . 3 7

Evaluation By f/ Date /[ 'I

//

NOTE: If these questions e e **yev'. the change must receiyu N.H.C. approval.

4. Unit Superintendent requests PO RC review O Check if YES.

              . 5. Approval

(' 34 % L , . te $w s,. ,,on s..r.< ne.,, _ ts ..r, _ e.nt

74/7? n. . 1438 208

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  . y.g. . . . q .:, .                                      .

2-33 tw-M.k %* e ., , Revision $3',@'.c - :.w - . x;.s . .. ' l.0 PURPOSE 0

                                                                                                                                                                                                                             ~
                       /~, ..                                                                                                                     .

t.,. ..--, (. This S0P ensures that all station sump discharges to the Industrial 1 .'c' 0..... . Waste Treatment System are monitored and sampled to ensure that 10

                                                                                                                                                                                                      .O Si t' '                                                CFR 20 MPC Values are not exceeded.                                                                                            5                                    'h gi-                                                                                                                                                                              R if;A.

5

   ,/     ,u.

2.0 LIMITS AND PRECAUTIONS N. ' .

                                                                                                                                                                                                 .,m ,
'$[: ;: .                                      2.1 The following Sump Pump Breakers will be maintained open unless g%.                     h.
                                                     , associated sump levels dictate pump operation. Prior to breaker                                                               Rw                 2. .A            N
         .d/g                                             closure and subsequer.t transfer of liquid to IWTS, a grab sample must                                                                                           .".

Dr be taken and an isotcpic analysis performed to ensure 10 CFR 20 MPC Table 2, Column 2 Values are not exceeded at the ' Station discharge.

jf@7 '

  %@g,D.C           o-
                                           .. .. In addition, permission to close sump pump breakers must be obtained                                                                                      .:p q 2
   'cr.g,' -                               .              from the Shift Supervisor. Caution Tags will be placed on each                                                                                                   ,
    ..,'                                     .          breaker referring to this SOP.                                                                                                                   y,:,

m *..m. . -

g. . .., <,'. ' Sump Sump Pump Breaker Location - '

     .hN                                      Unit I                  Turbine Room Sump                                                                                             (b.q d$J it.."

IWTS NOTE: SD-P-5 shall not SD-P-2A SD-P-28 ICTPMCC Unit 1C IDTPMCC Unit IE

                                                                                                                                                                                                           . . . . c; "

Mii. be used since it .

                                                                                                                                                                                                            .Fff
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bypasses the IWTS WM s.,,w

                                                                                                                                                                                                         . s+,w
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                                           ' Unit I                   Auxiliary Boiler                      SD-P-10A                          Local ut pump                                                ' :<.
  -mL                                                                 Blowdown Sump

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                      ';                                              IWTS                                  SD-P-10B                         Local at pump
  % i'                                        Unit I                  Powdex Sump                           SD-P-1A                           1ATPMCC Unit 4D                                        '..
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  +X                                          Unit 2                  Tur5ir.c Sldg. Semp                   SD ,                              2-31A Unit 3B                                          '-

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  '. .. .p>i.:2 Mj;.                                      IWTS                                  SD-                              2-41A Unit 9C                                          ' ^ i., , . '
  @h'.$.                                      Unit 2                  Tendon Gallery Sump                   SD-P-13A                      ~ 2-37 Unit' HG3                                      . M." _
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    .                         .                                       IWTS                                  SD-P-13B                          2-47 Unit JH2                 -
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Unit 2 Control & Service Bldg. SD-P-9A Sump 2-37 Unit ,EGl yl : . . j. u .

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S.I.d::r:. IWTS SD-P-9B 2-47 Unit GH2

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                                                                                                                                                                                           ~-]'?  ' .'.L        - l q: .

S&.i.%c~ Unit 2 Control Bldg. Area Sump SD-P-3A 2-31C Unit 4B G;R x:.9l," .' IWTS SD-P-3B 2-41C Unit SC ' -"?['g?- ~.

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9,: ,. m.- . IWrs SD-P-10B 2-llEC Unit 3CB a.< Q. .:=Mix ,- r

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                                                                                                                                                                     , ; q,yA,w.f..f.

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                                                                                                                                            ' 2-31 A Unit 10E           ,,..ys.M}D.t.;g                                       .     '
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                                                                 .                                                           . /. %               ,;<.y..                     u.;Pq.,

IMis .'.l.'?.' . . . Pretreatment MCC QF Un

Unit 1 Pretreatment Sump WT-P-24A . .ji[,.j%rm ...'#. .

                                                             .: -IWFS                                       WT-P-.24B         . . o .n.       Pretreatment MCC Unit 20...lnQ.2 WA Pi. -                  .
                                                                   .?     , . ,e   u: =n;-: ' . u - . .& ~

$gm.2..t.f+j:r ~.. ,m e. u.: w.L.. . ul... m..u . :... ..... w.w

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                                                                       ..                        .. . .                                                                                          .e.a.x 2:0   E.w        , ap      :wi::

w 1.4 38

Z-33

        'c.                                                                                                                                                                    ..<.t jg[.yt.,r   .;;k . . .'--
                                             ~

Revision $ J[.g '.

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, f h' - NOTE: Controls for Unit 1 Pretreatment Dua' Gravity Filter Backwa'rd .-: Flow, Skimmers, and Slu.'ge Collectors are not included in this l i

                ..                                            procedure since it could cause undue interruption of Pretreat:nent                                                                     \
   .,.T.l-t ..
 . :...w ..

System operation. These discharges are monitored at the IWTS Filtration Syste l. ..

                                                                                                                                                                       $0 oi,.    ,
   .y.y!y                               2.2 Immediately following sump pump-down open the associated breaker.                                                               ..
                                                                                                                                                                            . ?%:
                                                                                                                                                                           .i;.
.@w?w ~ -

2.,q. 3.0 PREREQUISITES ' - i . ..(

                                                                                                                                                                              .a..

7q.p: . . 3.1 One of the following sump levels is high and contents must be pumped to - ".3: s 9

    $:                                                                                                                                                                     ...t.1k
        .a .,.Q,:;-

the IWTS or IWFS. ., ,-

                                                                                                                                                                               ;s;...

5.9. _ y '. u. .W . Unit 1 to IWTS Unit 1 to IWFS. .

                                                                                                                                                                                    . ..c .

5 M;O.

 .g%             ;.                    .

Turbine Room Sump Unit 1 Pretreatment Sump h[ ' . . Auxiliary Boiler Blowdown Sump - Powdex Sump - qv.:. a .

                                                                                                                                                                            .n                   '

i ' Unit 2 to IWTS Unit 2 to IWFS *?" .

   & R.

Tj.@ Turbine Building Sump Unit 2 Pretreatment Sludge Collecti

 .yim Tendon Gallery Sump                                              Sump                                                                     ..
 ~@p.';';,                                       Control & Service Bldg. Sump                                                                                                     , ,/~ [ ~
               @                                 Control Bldg. Area Sump                                                                                                        ."..
                                                                                                                                                                           -. : ;? . ~

g'R s .,. ' Diesel A Sump g'~,y,Rev

          ?.w '                                  Diesel B Sump                                                           .

3.2 The sunp to be pumped down has had within the previous 24. hours an isotopic

,h. .

analysis performed on a sample of the contents and it is known not to contain lai concentrations.of radionuclides in excess of 10 CFR 20 MPC Table 2 Column 2 pf,D limitations takine 'nto account total plant effit.ent flow. ..g.. 2 ., w

                                                                                                                                                                                        ...I I6@.'                                          NOTE:        b cue s                 -

ndicates greater than MPC values of isotopes, M, the Shi.. ,iall review the DF and fraction of MPC after -

  $yp2.,~

(, - dilution calculations to . insure MPC is not exceeded at the plant - g;1., , discharge. Enclosures I & II. Sumps that must be pumped more ;. ~ hf.W frequently than every 24. hours shall be sampled at least every - Q$jf 24 hours. g.; .

                                                                                                                                                                                ..=..
 .I,&       ?y,                        3.3 Sump analysis results will be maintained by the Shift Foreman / Control                                                           Z' L's; t.y.';$..N          /                .         Room     Operator       in  the Water   Sample    Log    Book    in  Unit    I   Control          Room.                ,. 7.,.pu
                                                                                                                                                                           ^ . :c:.
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c.k: 4.0 PROCEDURE V .' ./: s!J.;dfy. t;&%.' - lf w 'P.::, .

 ~Mb   -                               4.1 Ensure Shift Foreman or Control Room Operator has obtained results of ' Y                                                            ' ' ' '
 .<.by.;                                         sump contents Isotopic Analysis and sum of the ratios of radionuclides WM-                                           is less than 1.0 at.the river.               Use IWTS/IWFS discharge flow rate (150 gpm)and Mw%

Effluent Flow Rate to determine dilution factor, Enclosures I & II.

                                                                                                                                                     - -               . m .... m->         .

If sump samples indicate greater than MPC, the Shift 1 Foreman .~Y'M-

   ,['er* ~ ,~. - NOTE:                                       must review dilution factor and fraction of MPC calculations to insure i'N: ...                                                     MPC is not exceeded at the plant discharge. This review must .$2 ? -

(,, be done prior to pumping the sump. ., j;,. s ;.: 4w.s. c,. j ;.jjg.;;t s,.

                                                                                                                                                               . . . .% .. . 4.w y7-e:.g,p;   e. .[:-              ,**.                                   ,-
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                                                                                                                                                                    ?55kk. .'-$5

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      .;                     -s                                                                                                                                                                                                 .
                                                                                                                                                                                                                                                      -                         .-                        7

( n(s. . , . c .g ,, fiOTE: The total fraction of f4PC (the fraction from this specific i k dW' sump and the existing fraction of f4PC discharge pt. 001 ,.c.. ., (plant discharge) must be less than one (1). See Z-51 ' ~' WW.* 't Rev.1 (line three of Liquid Release Form) and examples @( '" / . S.$..-,&.;.'6.

p. . at ECS station.

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                       .s.m;. i.:                        -
                                                                                                                                                                                                                                                                                                   .g..

M 4.2 Document less than f4PC at the station discharge by completing Enclosure II. ;r' %

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   ?g:'N.D,'.-    ;-                                                 '

w

                                                                                                                                                                                                                                                                      , ,. .p.a,                .'

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   . .                 M.. . .                                                                                                                                                                                                                                                                                    -
     ?.93;'?                                               4.3 Obtain permission from the Shift Supervisec to close the respective                                                                                                                              u 0;    . .i,.- , Ql3 '. .'
   .,p ? ."                                              .             sump pump breakers. Record the person granting permissir.i on                                                                                                                                "" M 3;'/.
   .2%p&                                                     . . Attachment I.                     flotify the flRC f?RR and I&E and the Radwaite groups                                                                                                          c .cy-',.dQ., ;. ,
                                                                                                                                                                                                                                                                        'MMY j $v

of intentions to pump sump and also record on Attachmat I.

e, .. - q fiOTE: De not use SD-P-5 to pump the Unit I tuvoine building sump. .f.h.c.. l v.q.$.. ~.c; . . . :;. ,

   "Ny       .                                             4.0 Close the sump pump breakers and allow the pumps to drau down the water                                                                                                                                                         j ':;'iX.

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ct. 4.5 Open the respective sump pump breaker. -

aLy.w. g, 4.6 flotify the Control Room (Tril-1) of start and stop times of sump pump- .u:.n{:;;:,:p.

                                                                                                                                                                                                                                                                                            .                 W d.y;' -;;,7                                                        downs and record.in both the Unit I CR0 Log and ECS Log.                                                                                                                                      -
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                         -c , .                            4.7 Attempt to ident fy and isolate the source and cause of all isotopic                                                                                                                                              ,.                        .

analysis high concentration indications.

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. ).; .s.. ,$$.\. .., . . E. -;Th#' I131 (pC.i/cc) = Concentration of 1 131 Found In Sample \' ' .>at.,.x. ' .ys ? .s- N'...
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N0TE: If any other isotopes other than I-131 are present in concentration > MDA the - ... weighted MPC shall be calculated and .-
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QUESTIO.1
56. Provide revised Figure 7-1 " Station Discharge" indicating sample points, in-line radiation monitors, and applicable unit numbers on the equipment.
RESPONSE
See revised Figure 7-1.
.1438 215
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QUESTION
57. Demonstrate conformance of the plant ventilation systems with NUREG-1.52 and NUREG-1.140 and provide Icak-tight testing procedures for the ventilation systems.
RESPONSE
Met-Ed currently is working with the Operating Reactors Branch of the NRC in order to finali::e Technical Specifications for the TMI-1 ventilation systems. The purpose of these new Specifications will be to invoke the testing requirements of Regulatory Guide 1.52 on these systems. In addition, before restart of TMI-1, the charcoal beds of the ventilation systems will be replaced with charcoal meeting the requirements of Regulatory Guide 1.52. After having accomplished these items the appropriate surveillance procedures will be availabic for NRC review. The affected ventilation systems are:
1. The Reactor Building Ventilation System (purge and exhaust) .

2. The Auxiliary and Fuel llandling Building Ventilation System.

3. The Control Room Ventilation System.
There are no systems to which Regulatory Guide 1.140 applies ,ince there are no normal filtered ventilation systems separate from the ai,ve systems. 1458 2\T
QUESTION SS. Does the Senior Vice-President - Met-Ed, Vice-President GPUSC have other responsibilities than TMI-1 and TMI-27 If so, describe the proportion of his time allocated to these other duties.
RESPONSE
The Senior Vice President - Met-Ed still has responsibil2ty for the operation of fossil plants belonging to Met-Ed and construction activities associated with new fossil plants; however, he is currently devoting his time to TMI-2 Recovery and TMI-l Restart Operations. 1438 218
QUESTION
59. Please clarify whether or not the position of Vice-President Nuclear Operations is a full-time onsite position.
RESPONSE
The Vice President - Met-Ed and Director of TMI-l (formerly referred to as Vice President - Nuclear Operations) is assigned onsite on a full-time basis. 1438 219
QUESTION
60. Describe the specific responsibilities any functional block of the Station Organization shown in Figure 5.3-1 has for TMI-2; and if so, the time normally allocated to these responsibilities.
RESPONSE
The only areas in which D1I-2 will be serviced by the DfI-l Station Organization will be in Training as indicated in Section 5.2.ll.f.2 and for those common functions under the Manager Support Services and Logistics, i.e., a D!I-2 Security Force and personnel administration. 1438 220
QUESTION
61. Figure 5.2-1 appears not to be consistent with the description on page 5-2 with regard to Rad. Chem. and !!.P. functions. Picase clarify.
RESPONSE
Figure 5.2.1 has been revised to reflect a change in organization and is consistent with the description on page 5.2 and in Section 5.2.10 1438 221
QUESTION
62. Sections 5.2.12 and 5.2.13, Station Organization and Maintenance Description, are shown in Figure 5.3-1 as Station Support Organization. Please clarify this apparent inconsistency. In addition, clarify the distinction between preventive and corrective maintenance.
RESPONSE
The Maintenance Organizations for Unit 1 and 2 are entirely separate. Sections 5.2.12 and 5.2.13 have been revised to reflect the current reporting relationships. Preventive Maintenance is defined as replacements, adjustments, major overhauls, inspections, and lubrications preplanned and scheduled on a cycle designated by engineering or by the maintenance departments in order to maintain equipment at normal efficiency. Corrective Maintenance is defined as work required to restore equipment because of failure of components either through lack of preventive maintenance or through design or performance inadequacies. 1438'222
QUESTION
63. Expand Figure 5.2-1 to show details down to the technician level.
Include qualification requirements for all functional blocks except those in the Manager Support Services and Logistics.
RESPONSE
See numbers in ( ) on attached figure, 1438 223
Figure 5.2.1 Senior Vice Pres. f4t-Ed Vice-Pres. GFU",C
' lice-President Me t-Ed Manager Manager Manager L27) flana9er (d}
t Hadiologicalg) Plan t Engineering t'n i t ' l g Controls ldninistration& ovviens f invor Emergency PORC Chairman Plannin9 Security [49), supv. Operator 3 I# U" LL) rne-dini+ne Training 2a o cg cal gentrols _ Lead Nuclear Superintendent ' Engineer hf f nagement o r s fiaintenance ,arta -at T a '
'^f*~'"
LeadElectricab jadinlogical Supv. Career E n c.i nee r Develop. Trng sentrols Supervisor OQ Supervisor (qq) - Surervisor o f(1) ' -oc8nici3, P re ven ti ve Corrective piirtrni-en f'peratinnt _p; t h i n *cna lead !&C E s u,,e rv i s o r Radiological q Engineer lead f*cchanical Supervisor (3) t Erqircerinq f o reman t'echa n i c al F r rn-in *aIWaata
, e a ,,
Technical Pechanical {}} Enqinane Cloctrical _ L e a'i Shift U) i staff foreman Electrical Surervisnr p'"' f In f"2" Shift icchnical { L4 M v 15 e r , I AC foremn ledd IEC Ferrman shift F o r eitu n Supervignr ll:- N Chemtgtrv fentrni RoomQ0i r r erator t e. h A n .41 y a t (F1rc l' r o t e r t 1 ,) LEGEND Alv i l i a ry (i f,] Q oparator g,ug,t;nn i(IN$@
QUESTION
64. Show the number of persons assigned to each of the functional blocks shown in the " revised" Figure 5.2-1.
RESPONSE
b'ithin the functional blocks shown on the revised Figure 5.2-1, personnel are assigned to each group as indicated below:
1. There are 30 positions authorized to be filled by the Manager -
Plant Engineering. Of these,30 are filled as follows:
a. Manager

b. PORC Chairman

c. 4 Lead Engineers

d. 6 Shift Technical Advisors

c. Supervisor Chemistry

f. 10 Engineers in various disciplines

g. 7 Chemistry Foremen / Technicians

2. There are 222 positions authorized to be filled by the Manager -
Unit 1. Of these, 167 are filled as follows:
a. Manager

b. Superintendent Maintenance

c. Supervisor Preventative Maintenance

d. Supervisor Corrective Maintenance

c. 3 Maintenance Foremen (PM)

f. 3 Lead Maintenance Foremen (CM)

g. 83 Bargaining Unit Personnel in the various disciplines

h. Supervisor of Operations

i. Supervisor of Radwaste

j. 4 Radwaste Engineer / Foremen

k. 6 Shift Supervisors

1. 6 Shift Foremen ra . 19 Control Room Operators

n. 36 Auxiliary Operators

3. There are 72 positions authorized to be filled by the Manager Administration and Services. Of these, all are filled with the exception of 4 individual positions in the personnel regruitment area, i438 225

4. There are 18 positions authorized to be filled by the Manager Training, exclusive of administrative support. Of these,11 are filled as follows:

a. Manager

b. Supervisor Operator Training

c. 4 Licensed Operator Training Instructors

d. 3 Unlicensed Operator Training Instructors

c. 2 General Employee Training Instructors

5. There are 34 positions authorized to be filled by the Radiological Controls Manager. Of these, 15 are currently filled by Company employees and 15 are filled by contractor personnel. The position of Radiological Controls Manager, Supervisor Radiological Engineering, and two Technical Staff positions are vacant. The position of Radiological Controls Manager is being filled on an acting basis by the Vice President -
Met-Ed. 1438 226
QUESTION
65. Clarify the meaning of c notation in Figure 5.2-1. If it is not a minimum requirement, delete it and any other reference such as
" preferred" that is not a minimum.
RESPONSE
Where indicated in a functional block, "C" indicates that a college degree is required. 1438 227
,QUESTION
66. Describe the function of each of the functional blocks reporting to the Director, Technical Support; Supervisor, Preventative Maintenance; and Superintendent, Radiological Controls and Chemistry.
RESPONSE
A. Reporting to the Manager Plant En ginee ring *
1. Chairman, Plant Operation Review Committee - Advises on all matters related to nuclear safety by reviewing all procedures and changes thereto, proposed changes and modifications to unit systems or equipment, and by per -
formance of special reviews, investigations or analyses, and reports.
2. Lead Engineer I6C - Reports on all matters relating to instrument and control systems and components including operating, test, and maintenance procedure and review, to ensure and p rovide sound engineering evaluations, test procedures, and maintenance recommendations nec- '
essary for the safe, efficient operation of the unit. Included are such systems as Reactor Protection, Integrated Control, Non-Nuclear Instrumentation, Incore Monitoring, Loose Parts Monitoring, pneumatic Control Valves and Componenets, Turbine Electro-Hydraulic Con-trol, Turbine Supervisory Instruments.
3. Lead Engineer Electrical - Reports on all matters re-lating to electrical systems and components, includine operating, test, and maintenance procedure and review, to ensure and provide sound engineering evaluations, test procedures, and maintenance recommendations necessary for the safe, efficient operation of the unit.
Included are such components and systems as Control rod drive, Pressurizer Heater Control Engineering Safeguards Actuation, Diesel Generators, Main and Auxiliary Trans-formers, Heat Trace, Main Generator Temperature Monitor-ing, Electrical Distribution, Grounding and Lighting Pro-tection, Bettery and Battery chargers, Inverters and Vital Busses, Electrical systems for Security System, and Substation equipment.
*formerly known as the Director - Technical Support i
1438 228 l
4. Lead Engineer Mechanical - Reports on all matters re-lating to the mechanical etgineering support of systems and components, including operating, test, and main-tenance procedure and review, to ensure and provide sound engineering evaluations, test procedures and maintenance recommendations necessary for the safe, efficient operation of the unit.
Included are such areas as Steam Generators, Reactor Coolant Pumps, Piping Hangers, supports, and snubbers, heat exchangers and coolers, Reactor Building integrates, containment isolation valve leak tightness, Emergency Diesel, ventilation systems, piping systems, valves, pumps, and filters,
5. Fire Protection - Reports on all matters relating to fire protection and associated systems and components, including operating, test, and maintenance procedure and review, to ensure and provide sound engineering evalua-tions, test procedures, and maintenance recommendations necessary for the safe, efficient operation of the unit.
Included are such areas as the Fire Service System, Penetration Sealing, Ventilation fire dampers, fire doors and walls, fire extinguishers and CO 2 nd Halon fire suppression systems.
6. Lead Engineer Nuclear - Reports on all matters relating to the operation of the nuclear reactor systems and components, including operating, test, and maintenance procedure and review, to ensure and provide sound engineering evaluations, test procedures, and mainten-ance recommendations necessary for the safe operation of the unit.
Included are components such as nuclear fuel, Peactor vessel internals, and fuel handling equipment. Techni-ca) support will be provided by recommending such core related parameters as shutdown margin calculations, estimated critical rod positions and transient Xenon consideration.
7. Supervisor Chemistry - Accountable for laboratory chemi-cal analysis, primary and secondary system chemistry control, water treatment, waste treatment, and radio-chemistry to meet established regulations and to ensure reliable plant operations.
1438 229
B. Reporting to the Superintendent of Maintenance
1. Supervisor of Preventative Maintenance - Plans, organizes, integrates, and directs the Preventative Maintenance pro-gram and the Technical Specification Surveillance program to ensure optimum equipment / systems utilization.

2. Supervisor of Corrective Maintenance - Directs the accom-plishment of all work required to restore equipment be-cause of failure of components or systems.

3. Lead Foremen reporting to the Supervisor of Corrective Maintenance direct the work of foremen and bargaining unit personnel in performing Corrective Maintenance in each of the three disciplines, electrical, mechanical, I6C.

4. Foremen reporting to the Supervisor of Preventative Main-tenance direct the work of bargaining unit personnel in performing Preventative Maintenance and Technical Speci-fication Surveillance in each of the three disciplines, electrical, mechanical, I&C.
1438 230
C. Reporting to the Radiological Controls Manager
1. Supervisor Radiological Controls - responsible for the coordination and performance of Radiological Control sup-port and enforcement functions accomplished by Radiologi-cal Control Technicians including contract technicians assigned to the Radiological Control Technicians Group.

2. Supervisor Radiological Engineering - resoonsible for all matters involving the Radiological Control program design and the technical aspects related to implementation of the program support functions including ALARA, Bioassay, add Respiratory Protection.

3. Radiological Control Foreman reporting to the Supervisor or Radiological Controls direct the work of Radiological Control Technicians who provide required radiological sup-port for station operators and maintenance activities and enforce Compliance with correct radiological control and work practices.

4. The Technical Staff reporting to the Supervisor of Radio-logical Engineering accomplished the design and technical aspects related to the implementation of the Radiological Control Program support functions including ALARA, Bioassay, and Respiratory Protection.
1438 231
QUESTION
67. There appears to be no commitment that the Superintendent, Radiation Centrols and Chemistry needs any qualification requirements in chemistry and radiochemistry. Please clarify.
RESPONSE
The title for tMs position has been changed to " Radiological Controls Manager". Thc ;umbent will meet the requirements of paragraph 4.4.4 ANSI /ANS-3.1 - - 78 and Reg. Guide 1.8 - 1978. The responsibility for chemistry will e assumed by a Supervisor of Chemistry who will meet ti.e requirements of paragraph 4.4.3 of ANSI /ANS-3.1 - 1978. 1438 232
QUESTION
68. In Section 5.2.15 you state that the Aux. Operator performs the function of radiation protection monitor. The Emergency Plan Section 4 states there will be a rad / chem technician assigned to each shift. Please clarify.
RESPONSE
The Auxiliary Operator performs the function of " radiation protection monitor on his shif t as required" as indicated in Section 5.2.15.b. This is not intended to be a primary responsibility since the function of radiation protection monitor is the responsibility of the Rad Chem Tech assigned to the shift.
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QUESTION
69. Describe the onsite fire protection organization including the reporting requirements for the Tech. Analyst SR-1.
RESPONSE
As indicated in the current revision to Figure 5.2.1, the Tech Analyst reports to the Manager Plant Engineering who is responsible for supervision of on-site fire protection. 1438 234
QUESTION
70. Describe the delegation of authority for overall station management in the event of unexpected contingencies of a temporary nature.
RESPONSE
The Vice President - Met-Ed has the responsibility for overall direction o f d ay-t o-d ay TMI-l operations including (directing the management staff identified on Figure 5.2-1 Station Organization), and Technical Specification and Regulatory Requirement Compliance. The Vice President - Met-Ed delegates the direct respon-sibility for operating the unit in a safe, reliable, and efficient manner to the Manager - Unit 1. The Manager - Unit 1 delegates the responsibility for the day-to-day administration and direction of the Operations personnel and ensuring Compliance with the conditions of the plant operating license and technical specifications to the Supervisor of Operations. The Shift Supervisor, who reports to the Supervisor of Operaticns, is the responsible management position on shift and as such shall manage and direct all activity in the physical plant and accepts full responsibility for all ac-tivity taking place on his shift. In the event of unexpected contingencies of a temporary nature the Shift Supervisor is responsible to take whateve' action is necessary to maintain the physical plant in a con-dition to protect and assure the health and safety of the public, the safety of personnel within the plant, and the safe operation of all plant systems and components. If the unexpected contingency expands into an event of a more serious nature, the Shift Supervisor is charged with actuating the Emergency Plan which requires specific notification requirements and specific mobilization actions. 1438 235
QUESTION
71. Provide greater detail on the personnel resumes already submitted (include relevent job assignments, education, and training, including dates for each). Include resumes of Lead Engineers reporting to the Director, Technical Support.
Response
The following information amplifies biographical descriptions for incumbents listed in Sections 5.2.4, Shift Supervisor, and 5.2.5, Shift Foreman: Incumbent A Education: High School Graduate - 1954 Military Service: U.S. Army 1958-1959 Relevant Assignments: Control Room Operator - 0ctober 1969 - October 1976 Shift Foreman - October 1976 - July 1979 Shift Supervisor July 1979 - Present Incumbent B Education: High School Graduate - 1961 Reievant Assignments: Auxiliary Operator - October 1969 - December 1972 Control Operator - December 1972 - July 1975 Shif t Foreman - July 1975 - July 1979 Shift Supervisor July 1979 - Present 1438 236
Incumbent C Education: High School Graduate - 1959 Military Service: U.S. Air Force - 1959-1963 Relevant Assignments: Control Room Operator - October 1963 - August 1976 Shift Foreman - August 1976 - April 1978 Shift Supervisor April 1978 - Present Incumbent D Education: High School Graduate - 1965 Military Service: U . S . flavy - 1966-1971 Relevant Assignments / Training: U.S. flavy Basic fluclear Power School (26 weeks) 1966-1967 fluclear Power Prototype School (26 weeks) - 1967 Reactor Operator - U.S. Navy - USS Bainbridge 1969-1971 Auxi'liary Operator - February 1972-October 1973 Shift Foreman - October 1973-Pay 1976 Shift Supervisor - May 1976-Present Incumbent E Education: High School Graduate - 1964 Military Service: U.S. Air Force - 1964-1969 Relevant Assign-: Auxiliary Operator - ments tiarch 1969-July 1970 1438 237
Control Room Operator - July 1973-October 1975 Shift Foreman - October 1975- October 1977 Shift Supervisor - October 1977 - Present Incumbent F Education: High School Graduate - 1963 Utah State University - 2 years Military Service U.S. Navy - 1966-1973 Relevant Assign-: U.S. Navy Basic Nuclear Power School ments Training (26 weeks) 1967 Nuclear Power Prototype School - (26 weeks) - 1967 Reactor Operator - USS Creenling - 1968-1973 Auxiliary Operator - April 1973-August 1975 Control Poom Operator - August 1975-November 1976 Shift Foreman - November 1976-March 1979 Shift Supervisor - March 1979-Present 1438 238
SHIFT FOREMAN Incumbent A Education: High School Graduate - 1963 Military Service: U.S. Air Force - 1963-1968 Relevant Assign-: Auxiliary Operator - ments July 1968 - June 1975 Control Room Operator - June 1975 - July 1978 Shift Foreman - July 1978 - Present Incumbent B Education: High School Graduate - 1963 Military Servicer U.S. Navy - 1964-1971 Relevan. Assign-: U.S. Navy Nuclear Power School - ments/ Training (26 weeks) - 1964-1965 Nuclear Power Prototype School - (26 weeks) - 1965 Mechanical Operator - USS Whale - 1968 Engine Room Supervisor - USS Theodore Roosevelt - 1969-1971 Auxiliary Operator - February 1971 - April 1975 Control Room Operator - April 1975 - May 1978 Shift Foreman - May 1978 - Present Incumbent U Education: High School Graduate - 1966 Military Servicer U.S. Navy 1968-1974 , 1438 239
Relevant Assign-: U.S. Navy Nuclear Power School - ments/ Training (26 weeks) - 1968 Nuclear Power Prototype School - (26 weeks) - 1968-1969 USS Sea Devil - Electrical System Operator - 1969-1972 USS Bates - Sound and Vibration Analysis - 1972-1974 Auxiliary Operator - February 1974 - October 1976 Control Room Occrator - October 1976 - July 1978 Shift Foreman - July 1978 - Present Incumbent D Education: High School Graduate - 1967 Relevant Assign-: Auxiliary Operator - ments January 1975 - October 1977 Control Room Operator - October 1977 - August 1978 Shift Foreman - August 1978 - Present Incumbent E Education: High School Graduate - 1957 Relevant Assign-: Auxiliary Operator - ments October 1969 - July 1970 Control Room Operator - July 1970 - June 1977 Shift Foreman - June 1977 - Present Incumbent F Education: High School Graduate - 1962, 1438 240
Military Service: U.S. Navy - 1963-1970 Relevant Assign-: U.S. Navy Nuclear Power School - ments/ Training (26 weeks) - 1965 Nuclear Power Prototype School - (26 weeks) - 1965-1966 Reactor Operator - USS George C. Marshall - October 1968 - March 1970 Auxiliary Operator - June 1970 - February 1974 Control Room Operator - February 1974 - February 1978 Shift Foreman - March 1978 - Present The following resumes are for Lead Engineers reporting to the Manager - Plant Engineering: I&C Incumbent Education: University of Louisville BSEE - 1966 University of South Carolina MSEE - 1968 Relevant Assign-: NSAEC Division of Naval Reactors (U.S. ments Navy - 03) - Nuclear Propulsion Engineer, I&C Section. Training - Bettis Reactor Engineering School; Reactor 6 months Design; 3 weeks - Prototype; 3 weeks - Shipyard - 1968-1973. Virginia Research, Inc. - Engineer 1973-1974 TMI - Lead I&C Engineer Training - Several Inst. v ndor courses 1 week B&W simulator 1974 - Present Nuclear In c utab e n t Education: Rensselaer Polytechnic Institute BS Nuclear Engineering - 1976 Relevant Assign-: Nuclear Engineering and Plant Performance ments Corporate Division Engineer I - Nuclear September 1976 - December 1976 1438 241
Nuclear Fuels - Corporate Division Engineer I - Nuclear December 1976 - January 1979 TMI - I Nuclear Engineering Engineer II - Nuclear January 1979 - Present Mechanical Incumbent Education: Pennsyl'ania State University BS Chemistry Military Service: U.S. Navy 1969-1974 Relevant Assign-: U.S. Naval Nuclear Power School and ments/ Training Prototype Training - 1969-1970 Qualified for Supervision of Operation and Maintenance of Naval Nuclear Propulsion Plant - 1970-1974 Two years experience as TMI-l Operations Department Engineering during first two years commercial operation of Unit and first refueling outage. Three years experience as TMI-l Lead Mechanical Engineer during commercial operation of Unit including three re-fueling outages. Electrical Incumbent Education: Pennsylvania State University Dubois Campus Associate Degree - Electrical 1963-1965 Pennsylvania State University Capitol Campus Eachelor of Engineering Technology - Electrical 1968-1970 Relevant Assign-: Summer Student - Engineering Assistant - ments TMI - June 1969 - September 1969 Project Engineer - TMI June 1970 - November 1973 Lead Engineer - TMI November 1973 - Present Holds Senior Reactor Operator License on TMI-I. 1438 242
QUESTION
72. Does the Director, Technical Function have any responsibility for any plant other than Th!I-1 and TF1I-2? If so, describe the extent of those responsibilities and the time normally allocr.ted to these responsibilities.
RESPONSE
The Director, Technical Functions is responsible for providing technical support to plants other than ThlI-1 as requested by the operating companies. At this time the technical resources devoted to stations other than Th11 are approximately 15% of the resources within the Technical Functions Group. 1A38 243
QUESTION
73. Please clarify ' he description of the .anctions of the Director-T51I-2 recovery in the fifth paragraph of Section 5.3. It appears to be in error.
RESPONSE
The description of the functions of the Director, ThlI-2 Recovery, in Section 5.3 is in error. The correct description is "The Director of TS1I-2 Recovery reports to the Sr. Vice President, b!ct-Ed/Vice President, GPUSC, and is responsible forthe effective and safe conduct of plant activities including operations, maintenance and modifications." 1438 244
.,a
QUESTION
74. Provide the number of persons assigned to each of the functional blocks shown in Figure 5.3-1.
RESPONSE
Current 1980 Plant Organizational Unit Personnel Personnel Systems Engineering 30 51 Engineering & Design 53 74 TMI Engineering Management 8 12 TMI-2 Recovery Engineering 4 20 Quality Assurance 43 53 The foregoing addresses those functional blocks which address the engineering and quality assurcnce resources available to support the TMI station. As described in Section 5.3.1, all staf4 except as noted under Systems Engineering, includes only GPUSC permanent personnel. Certain Met-Ed personnel are permanently assigned to the Systems Engineering Department. Support from outside contractors is not included in the staffing totals, but is available on short notice to supplement the GPUSC staff as necessary. Such support continues to be obtained, n r
QUESTION
75. Describe the qualification requirements in terms of education and experience backgrounds for each of the functional blocks shown in Figure 5.3-1 except for TMI-2 recovery. The breakdown should include cach discipline you consider necessary to providc support for the operations staff.
RESPONSE
The functional blocks considered necessary to provide technical and quality assurance support for the operations staff are as described in the response to Question 74. The operational quality assurance plan esta';1ishes minimum qualifications for technical management personnel as having a minimum of a BS degree in Engineering or Science, and a minimum of five years relevant nuclear experience. 1438 246 N 6%
QUESTION
76. Provide resumes of the managers and Icad engineers for each of the functional blocks shown in Figure 5.3-1 (except TMI-2 Recovery and Director - Environment, Health and Safety) and summary information regarding educational and experience background on the staffing of each of the functional blocks shown in this figure.
RESPONSE
For the organizational units providing engineering and quality assurance support to the station organization, the management incumbent possess at least a BS degree in Engineering or Science, and the following experience. Years of Eng. Years of Nuc. Title Exnerience Experience Director, Technical Functions 26 24 Manager, Systems Engineering 22 22 Manager, Engineering & Design 20 14 Manager, TMI-2 Recovery Engineering 14 14 Manager Quality Assurance 18 14 Summary information on the staffing of these organizational units is described in the response to question 74.
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QUESTION
77. Describe the specific responsibility, if any, of the functional blocks shown in Figure 5.3-1, other than Director TMI-2 Recovery, has for other plants than TMI-1 and the proportion of time they will be assigned to these other plants.
RESPONSE
With reference to the engineering and quality assurance support to the station organization, the specific responsibility of organizational units for support of TMI-2 is as follows: Systems Engineering Provides support to TMI-2 which is currently approximately 5% of total resources. Engineering & Design Provides support to TMI-2 which is currently approximately 15% of total resources. TMI Engineering Management Provides support to TMI-2 which is currently approximately 25% of total resources. TMI Recovery Engineering Provides full-time support to TMI-2. Quality Assurance Provides support to TMI-2 which is currently approximately 25% of total resources. 1438 248 e *' d e e
QUESTION
78. Is it intended that the station support organization be permanently assigned to the site?
RESPONSE
For engineering and quality assurance support to the station organization, only the T51I-2 Recovery Engineering Department is totally located at the station. Elements of the other organizational units are assigned permanent-ly to the station. 1438 249
QUESTION
79. Describe the regular assignment of all persons who will be assigned to the Emergency Organization shown in Figures 12 and 13 of Section 4.
RESPONSE
Ficure 12: Emerrency Orranization Accirment Current Assirnment Emergency Director Manager Unit 1 Technical Superintendent Unit 1 Supervisor Operations Unit 1 Co=municator Shift Technical Engineer (.The relief shift technical engineer will be utilized to fill this emergency position) Communications Accistants Various administrative positions Radiological Accessment Supervisor Radiological Engineering Coordinator Radiological Controls Engineer Supervisor Rad Waste Operations Support Center Operations Engineer Coordinator Administrator-Technical Training-Nuclear Group Supervisor-Technical Training-Huclear Radiological Analycic Support Group Supervisor-Tech. Training-Engineers nuclear Adn'lictrator-Tech. Training-Juc. Administrator-Tech. Training-Nuc. Health Phycica Coordinator Radiological Control Foreman Radiological Control Foreman Radiological Control Foreman Chemictry Coordinator Chemictry Supervisor Chemistry Foreman Cecurity "Ocrlinator Shift Sita Protection Serceant Site Security Force Shift Security Force Fire Brigade Team Shift Operations Personnel Monitorc (Variouc) Shift Operations & Rad-Con (Additional perconnel nay have Perconnel to be called in) 1458 250
Firure 12: Emercency Orranization Assirnment Current Assirnment Technical Support Center Coordinator Lead Mechanical Engineer Lead Electrical Engineer Lead I & C Engineer Manager Plant Engineering Technical Support Engineers Plant Engineering Staff Operations Coordinator Shift Supervisor Operations Coordinator Supervisor Operations (The on-shift shift supervisor , the Supv. Operations or another Shift Supv. designated by the Energency Director can fill this emergency position) Energency Maintenance Coordinator Shift Maintenance Foremen Emergency Repair Team (s) Shift Maintenance Technicians Firure 13: Enerrency Orranization Assirnment Current Assiennent Emergency Support Director Senior Vice President Met-Ed Vice President Met-Ed Assistant Emergency Support Director Vice President Met-Ed Unit 1 Recovery Manager Director Reliability Engineering Public Affairs Manager TMI Projects Communications Administrator I'ublic Information Coordinator Public Information ismergency Support Concunicator Supervisor Training-Jaclear epresent ative-Safety Administrator-Safety Group Leader Administrative Support Manager Administration & Services Director-Personnel Administration Exempt Supervisor-Administration 1438 251 - . .
Ficure 13: D errency Orranization Assiennent Current Assirnment Group Leader Security Support Manager-GPUSC Security Supervisor-Security Manager Corporate Services Group Leader Health Physics / Radiological Controls Manager Chemistry Support Manager Training Director Environmental Health and Safety Environmental Assessment Porter /Gertz Consultants Coordinator Engineer Assoc. Sr. II, S&L Personnel Monitoring Coordinator Dosimetry Supervisor Environmental Scientist II Group Leader Technical Support Director Technical Functions Manager Systems Engineering Control and Safety Analysis Mgr. Group Leader Maintenance Support Superintendent Maintenance Supervisor Lorrective Maintenance 1438 252 i
. . - - s _-. .--
QUESTION
80. Describe the extent of mobilization of your Emergency Organization shown in Figures 12 and 13 for each class of emergency; i.e. ,
(1) Unusual Events, (2) Alert, (3) Site Emergency, and (4) General Emergency.
RESPONSE
'I"ne extent of mobilization of the energency organization is detailed in Section 4.6.1 3 of the Restart Report. 1438 253 u ~ u .
QUESTION
81. The description of the responsibilities of the Manager, Support Services and Logistics, Section 4.5.1.1.d (page 4-50) appears to conflict with that described in Section 5. Please clarify.
RESPONSE
The Manager-Support Services and Logistics has been redesignated Manager-Administration and Services. The description of responsibilities have been revised in Sections 1.5.1.1.d 4 and Section 5 and are now consistent. 1438 254 _ .. _a
QUESTION
82. Describe the number of persons you plan to assign to each of the functional blocks shown in Figures 12 and 13.
RESPONSE
The duty section concept for staffing figure 12 is dicussed in the answer to Question 86. For the functional blocks where more than one individual is required,the annoted Figure 12 is attached. The number in ( ) indicates the average number of personnel expected to be assigned. The staffing of Figure 13 vill be accomplished by designating a primary and two alternate individuals to each of the key supervisory positions. For those functional blocks where more than one individual is required, the annotated Figure 13 is attached. The number in ( ) indicates the average number of personnel expected to be assigned.
)438 255
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QUESTION
83. Describe, for each of the functional blocks shown in Figures 12 and 13, the qualification requirements in terms of education background (discipline), experience, and training.
RESPONSE
Personnel assignments for the key positions in the emergency organisations of Figures 12 and 13 have been made such that these assignments are consistent with their normal position functions. Qualifications for current assignments are described in Section 5 1438 258
. . . . =
QUESTION
84. Describe any interfaces between the Onsite Emergency Organization shown in Figure 12 and the Offsite Emergency Organization shown in Figure 13.
RESPONSE
Interfaces between the Onsite and Offsite Er.ergency Organizations are deceribed in Sections h.5.1.3 2.a, 2.b, 2.e; h.5.1.h; and h.5.1.h 1, 4. and 5 1438 259 2
QUIISTION
85. Describe the maximum time it will take from notification of those persons assigned to the emergency organizations (Figures 12 and 13) until they have been mobili:cd, and are performing their assigned duties.
RESPONSE
Table 8 in the TMI Unit I Emercency Plan (Section h of the Restart Report) describes the manpover and timing considerations for mobilizing the emergency organizations. 1438 260 i
QUESTION
86. Describe the Duty Sections referred to in Section 4.6.1.3 (pages 4-87 and 4-88).
RESPONSE
There vill be three duty sectionc that will be on a " call out" rotation. The duty sections vill be staffed by TMI station personnel and vill be headed by a Duty Section Su.ierintendent. The assignment of duty raction personnel vill be formal, procedurized, and controlled such that a full complement of personnel is always available in case of emergencies. In addition, each individual on the duty section vill have a designated emergennv organization position consistent with staffing Figure 12 in the TMI Unit 1 Emergency Plan. This specific designation vill also appear in the duty section roster sheet. i / 45.1 1438 261
.- .s
QUESTION
87. Describe the training to be received by other than the station organization in regard to their duties in the Emergency Organization.
Note Section 4.8.11, page 4-134.
RESPONSE
Table 12 of the TMI Unit 1 Emergency Plan describes the scope and frequency of the training required for all TMI Generation Group personnel with specific emergency organization assignments. These personnel may be members of the station organization or the station support organization. Section h.8.1.1 has been revised to reflect this.
'{ h b0 .
2
QUESTION _ SS. Describe in greater detail your plans for the long-term recovery organization. This information should include the following: A. The maximum number of persons you plan to staff each functional block shown. B. The minimum qualification requirements in terms of disciplines (education background) and experience. C. The source of personnel to staff this organization. D. The proposed assignment of management personnel for each functional block. E. The time frame to activate this organization and have them functioning.
RESPONSE
A. The following numbers indicate the approximate range of staffing le. tels . It should be pointed out that actual staffing levels vill depend on the type of accident and scope of the recovery effort. Public/ Government ' . fairs - (6-12) Administration & Logistics - (3C-100) Task Management & Scheduling - (6-10)
'Iechnical Working Group - (10-12)
Industry Advisory Group - (30-60) Technical Support Group - (20-60) TMI-l Operations Group - normal station staff Waste Management Group - (10-30) Plant Modifications Group - (10-30) 1438 263
B. Title or Groun Qualifications Recovery Operations Manager Senior Vice President Met-Ed (Senior TMI Generation Group Officer) Public/ Government Affairs Company Officer Administration and Logistics Company Officer Industry Advisory Group Industry technical experts as selected S the President-GPU Technical Support Group Qualifications consistent with providing the services described in Section 5.3 TMI Unit 1 Operations Group Qualifications described in Section 52 Waste Management Group Key people vill be drawn from the Technical Functions Group Plant Modifications Group Consistent with the professional level required to meet design specifications , codes and standards C. Personnel to staff this organisation vill be drawn from the TMI Generation Group Station and Station Support Organization , consultants , other utilities, architect engineers, vendors, etc. D. President GFU - incumbent Recovery Operations Manager - Senior Vice President Met-Ed/Vice President GPUSC Public/ Government Affairs - Vice President Communications GPUSC Administration & Inciatics - Vice President Operations - Met-Ed Tank Management & Scheduling- Manager Management Services Technical Supnort Group - Director - TMI-2 TMI-l Operations Group - Vice President Met-Ed Waste Management Group - Director TMI-2 Site Operations Plant b'odifientions Groun - Director Pro.iects - GPUSC Technical Working Group - The above group leaders plus representatives fro 1 '!RC, ti.e Architect Engineer, the NS3S Supplier and the Industry Advisory Group 5
}430 ?b
r E. F etion h.5.1.5 indicates that a deliberate shift to a long term recovery organization vill occur if a complicated or long term recovery operation is anticipated. This shift vill occur when conditions have stablized. The offsite emergency support organization in Figure 13 is ,tructured to be compatible with the long term recovery organization. Once the decision is made to shift the organization into the long term recovery mode, the functional blocks in the Figure 13 organization can expand into the functional blocks in the long term organization (e.g. Public Affairs + Public/ Government Affairs , Group Leader Technical Support + Technical Support Group). The time frame to activate this organization, assuming that it is required, vill vary based on the time it takes plant conditions to stablize. There is no need to rapidly organize into this structure since the Figure 13 organization contains all the essential elements of offsite support and can be staffed in a time frame described in Table 8. 1438 265
QUESTION
89. Describe your provisions for keeping your emergency plans up to date, including assignments of personnel, their availability, and training programs, as necessary.
RESPONSE
Section h.8 describes the methods for updating the emergency plan and the training programs. Section h.5.1.3 describes the acci n .'t of personnel to the emergency organization. t 1438 266
QUESTION
90. Assure that your definition of " Safety Grade" in Section 1.4 of the Restart Report which identifies " redundancy" as a required feature, meets NRC single failure crit ria including separability.
RESPONSE
The term " redundancy" is intended to mean single failure proof and separation in accordance with the TMI-l FSAR separation criteria 4 1438 267 l
gtESTION J
91. The following additional information is required concerning NUREG-0578 item 2.1.5.B - Ilydrogen Recombiners:
A. You indicate that you have performed hydrogen generation rate calculation in accordance with Regulatory Guide 1.7 in order to verify that adequate time is availabic following an accident to install the second recombiner and still maintain the containment atmosphere hydrogen concentration with acceptable limits. Submit these calculations for our review. Indicate where the second hydrogen recombiner is to be stored prior to being used and verify that adequate procedures are availabic and training perfermed to assure that the second recombiner can be properly and expeditiously installed. B. You indicate that you will perform an evaluation to demonstrate that potential leakage and discharge to the atmosphere of the Intermediate Building air used for recombiner cooling will not result in off-site dose releases in excess of 10 CFR 100 limits. Submit this evaluation for our review.
RESPONSE
A. We have stated that the design basis for the system is a LOCA with hydrogen generation rates calculated in accordance with Reg. Guide No. 1.7. We have requested one of our consultants, Pickard, Lowe and Carrick, Inc., to evaluate and confirm that the recombiner which had been purchased for TMI Unit No. 2 is of sufficient capacity for TMI Unit No. 1. Note that the TMI-2 recombiner sizing is based on calculations in accordance with Reg. Guide No. 1.7 which indicate that the 3% by volume hydrogen concentration occurs approximately 250 hours after a LOCA. Since TMI-1 and 2 have approxi-mately the same available free containment space and they both have the same size reactors, the time at which containment hydrogen concentrations reach the 3% level are approximately the same. Based on the 250 hours at which time the recombiners are required to oper- e and the Standard Review Plan Section 6.2.5, Paragraph II.12e requirement, the stored recombiner must be made available to perform its function in a time period that is equal to or less than one-half of the 250 hours. Since the redundant recombiner will be stored in a seismic class I structure on the TMI site and all the connections for its use will be permanently installed, there seems to be no reason why the redundant recombiner cannot be made available to perform its function within 125 hours after a LOCA. A copy of the calculations performed by our consultant will be submitted to you as soon as we are in its receipt, which is expected by November 30, 1979. Adequate procedures for the maintenance and installation of the redundant recombiner will be made available to you as soon as they are conpleted. Training will be performed to assure that the second recombiner can be nroperly and expeditiously installed. 4 1438 268
B. The hydrogen recombiner system is an engineered safeguards welded piping
.ystem with the exception of the recombiner connection and flow element f.: anges . After installation of the redundant recombiner, the piping system shall be pneumatically tested to demonstrate that the leakage is within acceptabic limits in order to limit the off-site releases to below the 10CFR100 allowables. Since any possible leakage from the closed piping system will be insignificant compared to the total allowable post LOCA containment leakage which yields releases below the 10CFR100 allowables, we conclude that exhaust of the recombiner cooling air which would include the recombiner piping leaks would also be below the 10CFR100 limits.
t
;438 269
QUESTION
92. The Met-Ed response to Question 17 (Supplement 1 to the TMI-1 Restart Report) identifies loss of natural circulation due to loss of heat sink as a condition to be analyzed for inadequate core cooling response. This analysis meets the requirements of NUREG-0578 but does not correspond to the analysis (DNB transient at power) proposed at the September 13, 1979 B5W Owner's Group meeting. Please identify those analyses of inadequate core cooling which will be performed independently, or in verification, of the generic analyses to be provided by BGW.
RESPONSE
The e.ransient listed in our response to question 17 (loss of natural circulation due to loss of hint sink) was in error. This was intended to be consistent with the B&W transient to be analyzed (DNB transient at power). This listing has been corrected on the revised second page for the response to question 17, and a copy of this revision is attached. The RETRAN analyses which will be conducted by GPUSC/ Met-Ed are identified in Appendix A, Table 8A-1 of the Restart Report. s 1438 270
QUESTION
93. As requested by Question 18 (Supplement 1 to the TMI-1 Restart Report), please identify those emergency procedures which require the explicit use of (a) in-core thermocouples, (b) wide-range reactor outlet temperature measurement, (c) reactor coolant saturation temperature margin, and (c) other instrumentation identified in the response to Question 17 of the Supplement. Identify each instrument class separately.
RESPONSE
The following procedures have or are being revised to provide specific guidance on inadequate core cooling which require the use of: a) In-core Thermocouples EP 1202-6 Loss of Reactor Coolant /RC Pressure EP 1202-2/2A Station Blackout and Station Blackout With Loss of Both Diesels EP 1202-39 Inadequate Core Cooling b) Wide Range Reactor Outlet Temperature EP 1202-6 Loss of Reactor Coolant /RC Pressure EP 1202-26A Loss of Steam Generator Feed to Both OTSG's EP 1202-39 Inadequate Core Cooling Although specific reference is not made to the wide range reactor outlet temperature in all procedure the assumption that reference to the wide range indica-tion vill be made if narrow range is exceeded. c) Reactor Coolant Sat'; ration tieter EP 1202-4 Reactor Trip EP 1202-6 Loss of Reactor Coolant /RC Pressure EP 1202-26A Loss of Steam Generator Feed to Both OTSG's EP 1202-2/2A Station Blackout and Station Blackout With Loss of Both Diesels EP 1202-39 Inadequate Core Cooling , 1438 27I
OP 1102-16 f!atural Circulation EP 1202-14 Loss of RC Flow /RC Pump Trip d) Other existing Instrumentation EP 1202-4 Reactor Trip EP 1202-6 Loss of Peactor Coolant /RC Pressure EP 1202 14 Loss of RC Flow /RC Pump Trip EP 1202-26A Loss of Steam Generator Feed to Both OTSG's EP 1202-2/2A Station Blackout and Station Blackout With Loss of Both Diesels EP 1202-39 Inadequate Core Cooling OP 1102-16 Natural Circulation 1438 272
. :3
QUESTION
94. As requested by Question 20 (Supplement 1 to the TMI-1 Restart Report), please complete the description of the proposed subcooling meter. Include:
A. Overall display uncertainty B. Display qualification C. Information from page 2 of Tabic 20-1 (omitted from Submittal). If the device utilizes an annicg steam tabic approximation as stated, describe the process it. .iufficient detail to verify the applicability of the method. Include range and accuracy of the process.
RESPONSE
See revised tal;1e 20-1 of the response to Question 20 in Sdpplement 1, Part 1. 1438 273 ~
- . . .~b
TABLE 20-1 Information on the Subcooling Meter Plant Name: Three Mile Island, Unit No. 1 Vendor: Foxboro (Calculator) Reference for Information: Lisplay Information Display Tsat Margin, Psat Margin Display Type Dicital Continuous or on Demand Tsat margin continuous, Psat margin on demand Single or Redundant Display Single (Selectable) Location of Display Control Room Console Alarms _ Low Tsat margin (set point later) Overall uncertainty (Later) Range of Display 0-10000F, 0-1000 psi Qualifications (Not yet specified) Calculator Type Analog If process computor is used specify availability Not applicable Fingle or redundant calculators Redundant Selection Logic Highest T. Qualifications (Seismic, Environ, IEEE 279 Calculation Technique Steam Table Approximation For computing saturation temperature or pressure, the steam table satura-tion curve will be synthesized with 0.5% accuracy by means of a " Signal Characterizer" module. This is a solid state function generator which has the capability of simulating the characteristic curve of a process by means of a number of straight line segments. Up to eight segments may be used. The slope and intersection of each segment are individually adjustable.
)438 274 a
Input Temperature RTD's Temperature 4 sensors, T hot Range of Temperature sensors 1200 - 9200F Uncertainty
of temperature nensors (OF at 1) (Later)
Qualifications (seismic, environmental, IEEE 323) Control Grade (Short Term) Pressure (Specify instrument used) Pressure (number of sensors and locations) Two, PZR Press. Range of Pressure sensors 0-2500 PSIG Uncertainty
of pressure sensors (PSI at 1) (Later)
Qualifications (seismic, environmental, IEEE323) Seismic, Envit)nmental Backup Capability Availability of Temp & Press Individual press. 6 temp. indic. available in Control Room Availability of Steam Tables etc. Steam tables available on control panel and in computer Training of operators In accordance with operator retraining program Procedures Procedures will be generated for use of instruments Other Psat & Tsat will be independently computed by plant computer.
Uncertainties must address conditions of forced flow and natural circulation.
,438 ,2/c3 ? ,.~,e
QUESTION
95. Paragraph 2.1.3.b of NUREG-0578 requires a description of further measures and supporting analyses that will yield more direct indication of low reactor coolant level and inadequate core cooling such as reactor vessel water Icvel instrumentation. Section 2.1.1.6 of the Restart Report does not address further measures (to be impicmented by January 1, 1981), nor does it address the question of reactor vessel water level instrumes.*ation. Provide a conceptual description of what additional measures will be taken to detect inadequate core cooling. Provide an impicmentation schedule for these changes.
RESPONSE
Babcock 6 Wilcox is currently evaluating the need for further measures that will yield more direct indication of low reactor coolant 1cyc1 and inadequate core cooling. This evaluation will cover all the inadequate core cooling evaluation cases and is scheduled for completion ,1 December 14, 1979. Babcock 6 Wilcox has scheduled to provide recommendations for additional instrumentation (fi any) by February 1,1980. It is our intent to rely on this generic BGW evaluation. The conceptual description of additional measures to detect inadequate core cooling and an impicmentation schedule for any required changes will not be available until February, 1980, at the earliest. 1438 276 4,n.}}

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