ML19224C025

Plant Mods for Achieving Cold Shutdown May,1979 Prepared by Technical Review Group

ML19224C025
Person / Time
Site:	Crane
Issue date:	05/31/1979
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19224C021	List: ML19224C016 ML19224C025
References
SER-790531, NUDOCS 7906280176
Download: ML19224C025 (66)

Text

,

TMI-2 PLANT MODIFICATIONS FOR ACHIE"ING COLD SHUTDOWN

• MAY 1979

• This report was prepared based on infor-

=ation available to the staff prior to the plant being placed in a natural circulation mode of cooling on April 27, 1979. It is expected that certain plant changes different than those described herein may result and will be evaluated in a subsequent report.

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3 E11 -2 PLAf1T MOD I F I:ATIC'!S FCR ACHIE'!!'!G CCL3 SHUTDC'.fi Table of Contents

1 Background

2.

Steam Generator Modifications a.

Steam Generator "A" Modification - Short/Long Tern 1)

Design Conceot 2)

Modification 3)

Syste, Evaluatien 5.

S tean Genera tor "S Mod i f ica t i on - Short Te rn/Long Term 1)

Design Concept

2) Modification 3)

Systen Evaluacion c.

Mechanical Desic Evaluation (Stean Generator A/B) d.

Structural Evalu, tion (Steam Generator A/3) e.

Instrumentation

,d Control (Steam Generator A/B) f.

Radiological Eva aation (Steam Generator A/B) 3.

Reactor Coolant Syst, Pressure Control a.

Syste, Evaluation b.

Mechanical Design Evaluation c.

Structural Evaluation d.

Instrumentation and Control 4

Decay Heat Remo va '

a.

Upgrade of Evisting DHR Steam Leak Tightness b.

Skid Mounted DHR System 1)

Systen Evaluation

2) M2chanical Design Evaluation 3)

Sr.ructural Evaluation 4) instrumentation and Control 5.

Electrical System

di'ications 253 0D1 a.

General b.

Soecific System Modifications 7.

Qualitv Assurance a

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l.

B ackcround In order to bring IMI-2 to a cold shutdown condition, modifica dens will be made to various plant syste=s in phases to be carried cut over a period of the next few weeks.

These modifications will per=it a gradual transition from the current plant operating mode to ons which provides a stable long term cooldown mode of operatien.

The first planned modifications will be made in conjunction with the transitica from forced primary coolant circulation (bv che reactor coolant pump) to natural circulation through 1.he core.

To accomplish natural circulation, the secondary side (shell) of the stems generators will be operated water solid. Water will enter through the main feedwater rina and exit through the main see

.ine.

In order to provide for water solid operation, certain =edifications to each steam genera t or secondary flew loop will be required. The design of =odifications to ste rm generator "B" has accounted for possible contamination as a result of suspected tube leakage in the stems generator.

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. 2-Ta backuo cressure cont rol capability of the 7rinary system durine natural circulation, a new oressure control and rtkeup sys tem will be provided.

This system it essential in the event of loss of pressurizer heaters and level indication.

Criteria and procedures for letcown and overoressure control of the primary system will be established prior to going into this mode of operation.

Secause o' susoected leakage in the existing ciant decay heat removal system, a p rog ran, wi l l be conductec to identi'y and correct leaks to provide as leak tight a system as possible.

Also, an additional skid mounted decay neat removal train will be connected into the existing system as a cackup.

Connections will be provided to the new train for a cossible addition of a dedicated decay heat removal and cleanuo system Iccated in its own permanent structure.

To facilitiate early comoletion of design and installation of these system modifications, system functional capability following a seismic event has not Oeen a design requirement.

If a seismic event should occur and damage the modi #ied systems, the seismic Category 1 TMl-2 Decay Hea t Removal System and Reactor Coolant Makeup System could be used to remove core decay heat and control primary system pressure as necessary.

More detailed descriptions of these modi'ications are included in the follasing pages.

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'taa Gerarator Mndicat'nne

~ications to S*eam Ganerator "A" 'nr

.'a t e r Sol i ' Oca ra t ions I) Dasien Conceot The shor-and 1,ng tern concept for water solid coeration o' Steam Generator A have been consolidated.

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t'a n rs cuno :hrougn t"e s" ell side o' a new heat exchanger and into the secondarv. side o# the steam c.enerator A in a closed loco to remove heat 'ron the stea, generator.

The tube side of the new heat exchancer will be cooled bv the ex' sting Nuclear Services ?. i ve r date-System (NSRWS) which sunplies water from the river and returns it to the mechanical draft cooling tower.

Refer to Figure 1 for i schematic of this ' low path.

P-ovisions will be made for svstem pressure and expansion control av utilizing tne existing 3rd s tage feedvater heater she ll and its,itrogen suoply as a cressurized surge tank.

The des ign al so includes provisions for sanoli,-

de,i ne ra l i za t i on and chemical addition cacability.

For the initial phase of oceration, all valves, will be manually operated and instru entation will orovide local read out.

The new 1000 c' hea? re oval equioment has been designed to coerate at a acessure hic"er than the evoected eacto coolant system oressure thus assu-ing in-leakage o' secondav svstem n

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liquid into the crimary system in the event of stean generator tuce leakage.

2) Modification This scheme will involve installation of a nev high pressure train consisting o' a cuno, heat exchanger, va'"es and pipina lo_ated in

t. e turbina bui'di,n basement Th e loon vil! be connected to the main stea, turoine bvoass I;ne oetween the connection to the main steam lines and the condenser, and to the main feedwater line between the ~eedwater pumo and 3rd stage feedwater heater FW-J-6A.

Horizontal runs of piping will be succorted for static loads and secured to succortin structu es to prevent lateral motion.

Vertical runs of piping will be secured to cermanent structural members as required.

Additional oiping will be retuired 'o/ the su.~ge tank (3rd stage feedwater heater FW-J-6A), chemical oddition tank and demineralizer, in addition, the interconnections between the A and 5 fee dwa t e r heater trains will be broken and cacced off-Jumper cipes will be i ns tal l ed be tween,the ex i s t ing Nucl ea r Services River Vater System (NSkWS) the existing Seconda ry Services River Water Svstem (SSRWS) to provide cooling water to the tube side o# the new heat exchanger.

Th e safety classi'icat i:n of the Nuclear Services R i ve r '..'a t e r Svstem will be,aintained by oroviding double is]lation valves.

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. All piping connection < will be welded.

These modi fica t ions are excected to be comoleted and the syste? ready 'or coeration by t he m i dd l e o f lia y.

3)

Svste,s Evaluatinns The system design as croposed will meet the necessary secondarv side requirements for decay heat removal with either forced or natural circulation througn the reactor core.

All recuired modi #ications vill be made to acconolish this purpose.

This system is completely independent and separate from steam generator "S" during all intended modes of operation with the exception that the Nuclear Services River Water System and Secondary Services River Water System will be shared by both loops of steam generator cooling.

The system will not be provided with redundant active comoonerts.

However, a single active failure within the system will not comoromise natural circulation o# the primary system in that the secondary cooling loop through steam generator

'B" will continue to operate (see s ta f f evaluation of TMI-2 natural circulation performance).

The flowrates predicted through each of the heat exchangers will orovide adequate cooling cased on an assumed heat load 6

of 30 x 10 Stu/hr (RCP coerating + ] "WDH).

Coerating cerfor,ance and design parameters for the system are as

• ollows:

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6-Steam Generator A Modified System Goe ra t i no Condi t i nns Desien

Syste, Location Pressure (nsio)

Temo. (Or)

Flow (con)

Pressure (osio)

New Pump Discharge 670 100 3000-5000 800 New Puro Suction 500 100 3000-5000 600 (New Heat Exch.

Disch. Shell Side)

New Pumo Recirc.

670 100 Pur.; min. #10w DO

ew Heat Ex cna r.ge r 670 120 1000-5 C^0 Sucply (Shell Sice)

NSRW Supoly to New 100 85 6000 150 Heat Exchanger (Tube Side)

NSRW Return from New 100 95 6000 150 Heat Exchanger (Tube Side)

The systen Flow arrangement has been selected to minimize fouling effects bv maintaining Nuclear Services River Water on the tube side of the new heat exchanger.

Svstem operating temperature indicated #oe N S R'.

sucolv and return are design

values, b.

Modificatione t o S t aa, Ge ' ra t o r "B' f o r Va t e r Solid Onerations 1)

Desicn Conceot The short and long term concept for water solid coeration of Stean Generator 3 have been :,nsolidated.

One concent will be utilized.

later wi?1 be circulated tv :he new pumo through the tube side of the ew heat exchanger and ir.to t he seconda ry sice o' s tean generator S in a closed 1000 to renove hea! # rom the steam generator.

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. side (shell side) of this new heat exchanger will be cooled by the Secondary Services Closed Cooling dater (SSCW) system which is, in turn, cooled bv the Nuclear Services River Wreer Svstem, by way of the mechanical draft cool i ng towe r.

Refer to Figure 2 for a schematic o' this ficw cath.

Provisions will be made for system pressure and expantinn control av utilizing the shell side o# ne e<istina 3rd stage reedwater meater wi tn its ex'stino nitrnmen sucolv anc thus will 70erate as a cressurized surge tank.

The des i gn al so i ncl udes o rov i s i ons 'or samoling, demi ne ra l i za t i nn, and chemical addition capability.

For the initial chas? of coeration, all valves will be manually operated and instrumentation will provide local read out.

ine irst inte,ediate loop af this scheme (that portion of the secondary system which removes heat directly from the steam generator) has been designed to permit normal operation at a pressure higher than the reactor coolant system cressure, thus assuring inleakage of secondarv system liquid into the crimary system in the event of steam generator tube leakage.

However, the expected mode of operation from snich to initiate natural circulation would involve higher primary side

.; r e s s u r e s.

2) Mndi#icatinn This schecc.ill involve installation o' a new high pressure train consisting o# a pumo (LT3 P.1), beat exchanger (LT3-C-1 ), va l ves ano oicing located in t"e turbine b'silding casement.

The loco will be connected to the main steam line at a 10" drain act between the

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main steam isolation valve and the stop valve, and to the main f ee dwa t e r li.e between the #eedwa te r pumo and 3 rd s tage f ee dwa t e r heater FW-J-63.

Piping supports will be similar t:> those provided for S team Gene ra tor A mod i f i ca t ion <,.

Additional pioing will be required #ar the surge tank (3rd stage feedwa te r heat e r FW-J-6 8 ), chem i ca l a dd i t i u..

tant (L'.3-T-2), and de,ineralizer.

In addition, the inter-connections between the A and 3 feedwater heater trains will ce broken and cacped off Connections will be made to the existing secondary services closed cooling water system suoply and return lines for cooling the shell side of heat e xchange r LTB-C-1.

Jumper pipes will be installed between the existing Nuclear Services River Water System and the existing Secondary Services Piver Water Syste, to cool the tube side of the Secondary Services Closcd Cooling Water Svstem.

The safety classification of the Nuclear Services River Vater System will be maintained by providing double isolation valves.

All pining connections will be welded. These modifications are expected to be coroleted and the system ready for coeration by May 7, 1979.

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Svstems Evaluation The system design as proposed will meet the necessary secondary side requi rement s for decay heat removal with either forced or natural circulation through the reactor core.

All requ' red modi-

1. i ations will be made to accomolish this purpose.

Th i s sy s t em is indecen ent and s 20 ara te ' rom stea, generator 'A' ducirg all intended moces o f ope ra t ion wi th the exception that the Secondary Services R i ve r '.la t e r System will be shared by both locos of steam generator cooling.

The svst<.m uill not be provided with redundant active components.

H ow e v r. r, a single active failure within the system will not comoromise natural circulation of t he c rima ry sys tem in that the secondary cooling loco through steam generator "A" will continue to operate (see staf f evaluation concerning TMI-2 natural circulation per#ormance).

The fl owra tes predi cted through each of the heat exchangers will orovide 6

adequate cooling based on an assumed heat load of 30 x 10 Stu/hr (RCP operating + 3 MWDH).

Operating performance and design carameters for the system are as follows:

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Steam Generator B Modified System Oce-itinc Conditions Desinn Svsten Location Pressure (osic)

Temo (OF)

Flow (com)

Pressure losic)

New Puno D i scha rge 670 100 3000-5000 800 New Pumo Suction 500 100 3000-5000 600 (New Heat Exch. Disch.,

Tube Side)

New Pino Recire.

670 100 Pump min, flow 800 New Heat E-changer Supolv 670 127 3000 30:9 ogg (Tuce Side)

SSCW 5Jooly to New Hc a t 150 72 4000 Exchanger (Shell Side)

SSCW Return from New 100 59 5000 Heat Exchanger (Tube Side)

NSRW Return # rom SSCW 100 71 5000 Heat Exchanger (Tube Side)

The system flow arrangement has been sele ted to minimize fouling effects bv maintaining Nuclear Services River Water on the tube side of the Secondary Services Closed Conting Water heat exchanger _

0.

"echanical Svste, Dasion - Steam Generator

'A" and "B" Modifications All components and supports of both nuclear class and non-nuclear will be designed or verified to have been originally designed for the maximum loads that they could be exposed to during tes ting s tartuo, and exoected oceration of the system, i.e.,

pressure, temoerature ceacweignt, cumo vibration, etc.

The component design structural in'armat ion is listad in Table 1 L J.7') E )

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A scecific concern that we have addressed in our review is the structural adequacy of that portion of the Main Steam piping system which as in-corocrated into the OTSG cooling system will contain solid water in lieu of the pressurized steam for which i t was designed.

Components in the system will not experience any significant dynamic loads.

Special precautions will be taken during initial filling and startup of the system to minimize the potential for water hammer.

Loads that will be experienced include pressure, deadweight of water, and thermal exoansion.

Since the system will be operated at a maximum pressure of occut half :ne design pressure of the piping and its maximum ooerating temperatures will be considerably l owe r chan the design temocrature of the main steam ciping; stresses resulting from these loads will be minimal.

A'ter assemoly and prior to initial operation of the plant, the existing piping was hydrostat ic tested and at that time was water filled.

Thus the cioing and its supports have been demonstrated to be adequate for the weignt of the water.

In order to minimize piping deflection, the licensee has specified that selected spring hangers be pinned.

'Je concur wi t h this requirement.

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. Desien Bases Loads It should be noted that all ASME Section ::1 CL. 2 components used in the cooling system were designed for seismic Catec'ery I service.

Haw-ever all of these components, both those that are part of the original TMI-2 Main Steam and Feedwater piping sys tem and those obtained from other nuclear sites to be incorocrated into the 0T50 cooling systen, are being utilized in a system with different re.

onse characteristics from that for which they were ini tially des igned or a re operating with a fluid media different ' rom that for which they were seismically quali-

fled, i.e., some components designed for operation on steam during a seismic event as opposed to water filled as in the present system.

Thus because of these di f ferences f rom

• e original seismic design requirements, which can affect seismic resoonse, these components should not be considered seismically qualified as installed in the proposed cooling system, soiely on the basis of their original qualifications.

Additional work would be recuired to evaluate the seismic capability of these components for this aoplication.

However, seismic capability of these system modifications is not a necessary acceptance criterion; therefore, no additional seismic evalcation of this system is planned.

Evaluation Conclusion We have concluded that the Licensee has scecified ccmoonents designed and #abricated in accordance with acceotable industry codes or stan-darcs and will take into account the loads associated with startuo, testirg, and the planned system operation.

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. in conforaance with these cri teria The use of comoanents that are orovie, adequate as trance that structural integrity of the OTSG "A'

and "B" cooling system will be maintained.

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. Ta b l e 1 Comocnent Desian Structural Information Pumos A & B Svstem ASME Section !!! CL. 2 Detigr T

r.n ure - 350 F Design Pressure - 70' osig.

Heat EMChancees A Sys a,

ASME Section Vlli Design Pressure:

150 psio (tube side) 600 psig (shell side)

B Svsta, ASME Section ll1 CL. 2 Design Temne ra tu re :

350cF (Both Shell and Tube Sides)

Design Pressure:

675 osig ( ube side) 200 osig (shell side) 1rd Stace Feedwater Heater (shell side as succe tank)

ASME Section Vill Design Pressure:

1000 psig Stea, Generator Secondary Side - ASME Section ill CL. 2 Pioinc. Valvas and Mis. Tanks (minimum requirements)

Picing - ANSI 331.1 Misc. Tenks - ASME Section Vill Div. I Vai se s - A?!51 816.5 and 316.34 ree c eev,ater check ealvas outside coatsi-ent in ?"o -ain crea, isola!!,n va l ra<

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Structural Jesien - Steam Gena-ator

A" and "9" Mndifications Modification of Steam Generato "A" and Steam Genera tor "S" in as installation o' one pumo and one heat evenanger for each steam generator.

The new equipment will be installed

.n the north-west end of turbine building at e!evation 2 l'-8",

between c amn lines TJ and TK in nortn-south direction and TL1 and TLL in east-~cs: direction.

Tr e encios : rigure 1 snoas the general area of 1:ca::,n 0

'he ancee equiena,t.

Tne cetails of the ecuioment is as f o l l ows :

The heat excnanger for s team generator

'3" ne i gns 35 kips wet and is succortec on two saddles, 2 ' -G" b v h ' -0" e a n.

The au o for this Steam generator weigns 13.7 kips.

The neat exchancer #cr steam generator "A" weicns 135.1 kips we:

It is a=croximate1/ 33

't long and is succorted on two saddles 0 in w i d e a n c 4 ' long located 13 '-l' apa r t -

E x a.,i n a t i o n of the existing structural crawings o' the turbine building area, where 5is ecuionent is to be installed, reveals that the base slab is 'our 'eet thick, with 111 rebars at 12 in. scacing, each way,

00 and botto.,.

Too a' Se structural concrete base ma: is at the el" 271'-T' The base -at is covered with a 3'0" Tayer o' lean concrete 'ill p lus aco ro < ima tei-in, wearirg sur' ace, 7einging the

oc elev. to 2'I' Assu-ing L5 degree dis: ::utior t'e load,

-d on :%e casis o# :Se a'!?aaole cor: rete e--inc, essure, :Se za 7.

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. licensee estimated that the maximum load tnat can ce olaced over one square foot of the area is 30 kios.

Since the heaviest piece of equipment to be installed in this area is the heat excnanger, 135.9 kios, which when distributed over the area of the supporting saddles

.67' - 6.1 sc. 't) delivers the bead to the ' ', o 'r o

(2 x h.67' r

x

2 '< ins, the lic:. wee concluded t*at the s!ch is ccnaale to suoonrt the load.

The analvsis was per#ormed 'or the dead and live loacs on!v, uncer static conditions.

'.Je have concluded that the licensee has performed his analysis in accordance with the methods and crocedures which a re scecified bv the aporcariate codes a,a standards.

The use of these e thods provide a reasonable assurance that the structural alements affected by this modification will perform their intended function.

e.

Instru-eatation and Control (O TSG "A' and "B" CO,l i n el The following new instrurentation (l i s ted bel ow) is being provided on the OTSG "A" and "3" Long Te rn Cool i ng Sys tems.

1, o.adiation Monitor (*

'S' only) 2.

Loco f l ow Indication

• j.

Steam Generator Heat exchanger, Shell Side Te perature - T;n

(*A oni v), Tout Tn, Tout 5

Heat exchancer, Tube S ide Temoera ture -

i 5.

Pumo Pressure - S;ction a,d Jischarce

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Loco Temoera*u et E

Loco Pressure Hecundant sets o# instrumentation are marked.vith an atterisk.

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i Control room alarns will be orovided for radiation as a result of high S/G out-leakage.

For initial installation and system oper6 tion, only local indication will be available mounted near the sensors in the turbine building.

Similarly, all controls will be loca' S/G 1eakage into tne cooling loops is to be sensed by victoreen made:

355 area moni tors straoped to the piping just down stream of the tie-in to the.-ain steam line.

Loop flow rate is to be sensed by 2 (two) carton model 200 differential pressure mechanical indicators across a single permutit orifice olate.

All new temperature sensors / indicators are speci'ied to be ashcroft (5 - inch code 50 EI) bi-metal, liquid-filled thermometers.

Loop pressures are to be sensed by ashcroft 1279 (bourdon system) mechanical pre -

.e gauges.

Table i provides additional instrumentation characteristics.

The system desicn criteria include the requirement to provide control and sensor read-out to the main control room on an expedited basis.

Due to the time constraints placed upon initial system operation, we find the above design criteria to be acceptable. The specific details of the design associated with providing control and sensor read-out to the centrol room have not been developed at this time.

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'4 2 O TSG "tr' C001. IllG LOOP - TAbl E I VARIAHLE SENSOR TYPE LOCAT10ll S/G "ll" teakage Strap-on gamma Detectors I monitor on existing main steam turbine (primary to seconJary)

Victorcen bypass heatler ("A" loop only) 2 monitors on new, pipe just downstream of ex i s t ing sia i n steam header

('B l oop on l y)

Loop flow Total Flow Orifice Plate-Permutit 2 I ruli c a t o r s in new pipe just downstream of ex i s t i ng ma in s t ean header ("A" lonp) 2 indicators in new pipe Just upstream of existinq cedwater header (" B" loop)

D isne t a l l i c in new piping juct off mainsteani header Thermometer-Ashcroft a) llea t Exchanger Shell Side, Tempe ra t u r e Dimetallic At heat exchanger Thermometer-Ashcroft T in, To., r 8

tallic At heat exchanger Tobe Side, Teinp e r a t u r e Thermometer-Ashcroft Tin, T out Pump Pressure Suction and Discharge t'rcs sure Gauges-Ashc ro f t At discharge and suctlon of pump Loop Temperature Bimetallic Downstream of pump recirc.

II.e rmome t e r s-f' hc rof t u

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Loop Pressure Picssure Cauge-Ashcroft Upstream of heat exchanger CO r.:

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I f.

Radiolacical Evaluation (Steam Ge ne ra t or A/B)

The radiological considerations which are integra' tc the water solid steam generator secondary side cooling method will assure that radioact ive ef flue.its f rom contaminated sys tens will be controlled and

,inimiced.

Additional precuations will also be made to mini?.:za occupa-tional -adiation exposures to operating personnel.

The secondary coolant presently in the "3"

steam gener.t ' is contaminated due to the initial primary to secondary leakage which occurred on March 23, 1979.

The measured radioactivity concentration and curie content is estimated in Ta b l e 1.

Under normal conditions tne secondary coolant pressure will be ma into.ned a t a value greater than the crimary system pressure such that if steam generator leakage flow oaths are available, the highly contaminated p r ima ry coolant will not enter the secondary coolant.

However, it is expected that t rans i en t s of short duration may occur sech that a reverse pressure gradient could introduce additional radicactivity into the secondary coolant.

To alert the sys tem operators of such a condi tion, i nd ica to rs and ala rr.

for pressure and radioactivity in the seccndary coolant have been provided.

These ir.dicators will alert the operator of an adverse condition so that corrective action can be taken prior to signi#icant additional contamination of the secondarv coolant.

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g'iN..'# W~m,5 2 '\\'c'ac e-JNcT 'd, 'J", ? W1. 6 x l The s tea g3nera tor secondary coclant system and secondary service-closed cooling water system will be per;odically sanoled and analyzed to determine if heat exchangers are starting to leak.

Samples will be taken at a frequency of at least weekly or at any time there are indications that possible leakage may be occurring, e.g.,

in raase in s

tne steam generator closed cooling loop surge tank levels.

Leakage nav also occur at various mechanical connections in the secondary cooling system.

To the extent practical, locations where leakage is likely, e.g.,

valves wi*h a n own leakage history will be evaluated and a program to mininize lea ige implemented.

Even with a leak minimization program leakage of to :aminated liquids to the fl oo r drain system may still occur.

This lea age should f l ow into the floor drain system and be collected in the turbine building sump, the turbine building sumo cumps will be averated in a manual mode with an ana'y:is o' the sump wcter radioactivity content aeing made nrior to sump pump coeration.

If the radioactivity content i s l ow',

the water can be dischargac through its normal f l ow path to the river.

l' the radioacti vi ty content i, high, such that discharge would exceed technical soecification limits, the sumo water wil' be pumped to an appropriate rzdwaste system for treatment.

Adequate free volume in t*e radwaste system will

'e provided o

for such contingencies.

253 0: L

l

. Isolation between the secondary cooling system and services systems, e.g.,

nitrogen and demineralized water supplies to the surge tank, will be provided to prevent back fl ow of contaminated water.

Gaseous ef fluents fr;m thic sy, tem should be negligible.

The noble gas inventory in the "B" s team gene rator is negligible because t'e steam generator was vented.

Airoorne radiciodine releases should also be negligible because the secondary cooling system is not vented (a nitrogen blanketed surge tank) and the low secondary cooling svstem te,oerature (1000F) resul ts in a low air / water partition factor which reduces the volatility of the radiciodine.

The licensee is providing a demineralizer system to maintain secondary water chemistry and to reduce radioactive contamination.

Details on this svatem are not available at this time.

The demineralizers will be shielded and designed to permit

.,in bed disposal as radioactive waste.

The shielding and spent resin handllng syste.m will also be designed t:

minimize occupational radiation exoosures, e.g.,

the use of disposable dem i ne ra l i z e r s.

Folloaing the TMl-2 incident there has been no indication of primary coolant leakage cast t he

A" s team gene ra t or tubes.

Since the crimary to secondary sy, tem pressure di f ferential will result in in-leakage only, precautions similar to thor.e evaluated scove for the "B" steam generator modification are not recuired.

2b3 02?

Table 1 "3" Steam Generator Radioactivity Content (Ac ivities based on 4/19/79 sample reported by S&W)

April 19: 1979 May 9, 1979 Halt-Rastoacttvity (i>

aaatoace2.luf (3j

(;;

Life concentration Systen concentcation Systen (uC1)

Invento ry (uCi)

Inventory gm (C1) gm (Ci)

I-131 8.05d

.93 50.8 8.9 x 10-2 n,9

-3 Cs-134 2.ly 9 x 10

.59 5.3 x 10~

0.59 Cs-136 13d 8.x x 10~

.54 1.6 x 10~

0.18 2

-2 Cs-137 30y 3.4 x 10 2.2 2.0 x 10 2.2 7

(1)

S.C. water level = 358 inches = 6.5 x 10 cc (2)

S.G.

full at 525.5 inches = 1.1 x 10 cc (3) Assunes decay from 4-19 (21 days) plus dilution from filling the steam generator.

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I i 3.0 STANO3Y REACTC9 COOLANT PRESSURE CONTROL SYS TEM A.

5"sta Evaluation

1) Descriction A standby reactor coolant pressure control and makeup system has been proot sed bv the licensee.

TSis system would serve as a backuo to the

^'!CS anc aintain reactor coolant system pressure with the cressurizer #illed solid with water.

Primary coolant system pressure will be maintained even with the loss of pressurizer instrumentation and inoperative pressurizer heaters. Also, the pressure control syste will be designed to provide adequate NPSH to the reactor coolant pumos if they are needed.

The t tanby reactor coolant pressure control system will consists of passive components (a series of water storage tanks and a surgetank with nitrogen blanket for pressure control) and active components (cha rg i ng pumo s ).

The system will control reactor coolant pressure over the range of 100 psig to 750 psig.

The passive reactor coolant pressure system wN:-h wou'd be operatec n.tially with local control. Additional instrumentation and remote control will be incorporated to permit automatic operation of the system.

The active pressure control portion would resupply water to the surge tanks with acded cacability of providing additional makeup water direct!y to the RCS if needed.

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2) Modifications The s tandby reactor coolant system pressure control and makeup system will involve installation of two 900 gallon capacity water tanks and one surge tank (all will be of the Westinghouse Boron Injection Tank design), nitrogere bottles, two 49 gpm positive displacement oumos, a degassed barated wate r sucolv tank, valves, and oicing.

This system will be conrected between the di< charge side of the hign pressure makeuo system d owns t r eam of valve MU-V-Ik4C and upstream of valve MU-V16C.

All the components will be placed in the fuel handling building.

The above modi fication will establish a flowcath of makeup water and pressure control through the normal makeup lines that connect with the reactor coolant loop cold legs.

Chemical control of the degassified borated water used in the pressure control system will be provided by the p esent chemical addition system.

Connections will be provided to accommodate the addition of boric acid, H2, demineralized water and hydrazine, LiOH ano NaOH.

The degassed water tank will be replenished via pioing connection from borated water transfer pump and boric acid batching tank.

The boron concentration will be maintained in the range of 2200 to 4000 ppm.

Figure 1 depicts the oroposed pressure control system and interface connections to the existing systems.

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Evaluation

!f required, the passive portion of the standby pressure control and makeuo system is designed to provide peak initial 500 gpm injection rate to the reactor coolant system.

The 500 gpm injection rate will be adequate to provide primary system makeup for certain transient events that can cause considerable shrinkage in the 3CS.

3ecause of the finite inventory (1300-2000 gal) this injection will decrease as the discharge proceeds.

Also, the passive portion of the system will be designed to provide continuous makeup of k gom for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Su'ficient makeup volure requirement can be met by the proposed RCS pressure control system for moderate sys tem certurbat ions and for the following postulated transient event:

loss of natural circulation cooling due to a loss of all secondary side cooling with restart of one secondary cooling loop fo!!owing a hot leg temperature rise of 50 F.

For this event the licensee has shown a total volume change of 1900 gallons which can be made up by the proposed system.

However, the pressure control system is not designed for makeup requirements of more severe transients such as a sudden complete loss of natu a!

circulation for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, followed by an RC pump start.

Procedures which permit reactor coolant pumo restart following loss of natural circulation would require t.c avai ability of other makeuo svatens such as the HPI in additio, to the proco'ed pressure control System.

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. The reactor coolant pressure control and 9akeup syste n will not provide letdown cacability of reactor coolant caused bv overpressurization events.

Reactor coolant fluid exoansion will be relieved by one or a combination of the 'cIl owing cur rent components and systems:

(1 )

normal l e t d own line (through the letd&an coolers): (2) mainta ining letdown with concurrent ter,i atio, o# nakeuo/ seal i nj e c t i on. (1) coati'aed reactor coolant

cro seal return 'l w; (4) opening o' p res s urize r vent valve; and (5) li' ting of the pressurizer rel ie f valve.

Piping integrity of the reactor coolant oressure c,ntrol and makeuo system has been examined for postulated overpressurization events such as inadvertent startup o' an HPI puep.

Thi s sys tem wi l l nave a design pressure of 10)0 psig except for the section ' rom the HPI makeup line bat through the second isolation check valve which will have a design pressure o' 1500 psig.

Overoressurization protection of the latter pioire section when the HPl/makeuo pumo is started w'll be orovided by installation o' a relic # valve about the HPl/-akeuo cumo set at 1000 osig and a relief ' low ate of ':S gpm.

Check valves located downs t rea, of the HP!/ makeup pumos that are inside the Reactor Building will also provide protection to this system from inadvertent overpressurization of the RCS due to any other causes.

The criteria and crocedures 'or letdown and overpressure cont ol o# the 305 will be established prior to the oceration of this svstem.

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The proposed reactor coolant pressure control and makeup system may not meet makeuo demand following a depressurization event of the crimary system such as inadvertent opening of the pressurizer PORV (wi th downs t ream isolation valve openi.

If not isolated in time, the system would be draineo resulting i, entrainment of nitrogen into the reactor coolant system.

To orevent such an occurrence the system will automatically isolate on l ow level in the water tank.

In the event of loss o# o ##s i te powe r this valve wili fail in the "as is" position.

Since this position is the preferred position for normal operation and it would also be the preferred position in the event of loss of o'Fs i te power.

Also, an alarm will be annunciated locally and in the control room when the isolation valve is not fully open.

We have exar ined for single failures that can disable the pressure control and makeuo system.

The discharge valve SPC-V5 is a single fa:Ture point, h owe v e r it will be a -anually operated ball valve positioned to an open oosition, and then locked in that position.

For simplicity of design and installation, we have not required redundant valves to meet single criterion to insure system isolation capability. A redundant charging oumo will be available to fill the water tanks in response to tank level reduction.

.)

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s 29 _

• st of the RCS pressure control Requirements for oreoperational f l ow

.e system will be determined pending a review of the ma rgi n suggested by analysis for peak makeup flow demand and what the system is designed to provide.

conclude that the proposed system wiff aintain reactor coolant oressere control for nor-al water solid conditions and provide sufficient makeuo water for a wide range of evpected transien-events that would cause shrinkage in the RCS.

Emergency procedt.re: will be u ed for additional makeuo capability that may be raqJired to nitigate more severe and less probable transient events.

B.

Mechanical Da<ien 1)

Descrio,tjz The Reactor Coolant Pressure C7ntrol Systen is nadeur of several water supplv tanks, positive displacement pumps, valves, and piping.

The avolicable desicn codes and standards used for the design af these components are provided in the Table 1.

n.

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. Ta b l e 1 Acolicable Desien Codes or Standards Water Sunolv Tank (Passive Svstem)

ASME Section ill CL. 1 - S.S.

Cha r,i no P,os ASME Section 11! lL.2 Finina ANSI B31.1 (.:i; n imum requ i r emen t )

Decassified Vater Sucolv Ta n_k ASME Section VILt Div. I Comoonent Succorts ANSI B31.1

[3.)

ddV

=

O

I 2)

Jesien loads Pressure, pump vibration, cceponent and fluid deadweight, and maximum anticipated pressure surge forces were considerrd in developing design loads for the RCS makeup system.

The staff considers these to be acceotable load considerations for this coplication.

The staf f has noc required the RCS pressure control and akeus svstem be evaluated or s e i s.m i c load capability.

3) Other Considerations Welded construction will be used wherever possible to minimize the ootential for system leakage.

Component s will be f abricated f rom stainless steel or carbon steel clad with stainless steel.

4)

Evaluation Conclusion We conclude that the licensee has soecified components that will be designed and fabricated in p ordance with accentable industry codes or standards and will taken into account the loads associated with startuo, testing, and expected system operation.

The use of components that are in conformance with these cirteria provices adequate assurance that structural integrity of the Reactor Coolant P essure Control System will be maintained.

') t ' 1 is LJJ U),

6

1 iural Desien C.

O the eys et includes the follow'r.g major equipment i)

Three water tanks, 900 gal, cacacity.

Each weigh 20 kips emccy and 27.$ kips when full of water.

Each tank will be supported on four 12 in. by 12 in. plates.

nn four 6 in. plates.

b Scrated water tank weigne,g 62 f<ips supported c)

Two - 100 HP pumps weighing 5.L kips each.

The enclosed Figure 2 (2 sba-*-)

show the conditions of the original structure.

The area is loca ted between columns AP and AT in south-north di rection and ca r:nns A65 and A67 in east -west direction.

The slab is three feet thick and the reinforcing steel is ~3 at 9 in. too and bottom in north-south direction and #8 at 6 in. top and bottom in the east-west direction.

The compressive strength of the concrete is 3000 psi and the yield stress of the reinforcing steel is 60 ksi.

The licensee has analyzed the slac for the additional loads resulting from the new equip-er.t and concluded that the stresses will be within the allowables using generally accepted codes.

The analysis was perfor,ed for the static conditions only.

We conclude that the licensee has cer'or ed his analysis in accordance with the methods and procedures which are soecified by the accrocriate codes and standarcs.

The use o' these methode orovide reasonable assurance that the structures a##ected by this modi #ication will contir.ue to cer#orm their intended

' unction.

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In-tru-entation and Control The following instrumentation is to be provided for the Reactor Coolant Pressure Control System:

1.

System Water Pressure 2.

Water Level in each tank 3

Nitrogen Pressure 4

Make-up Flow 5

Make-up Pui.o Discharge Pressure c.

Borated Water Storaga Tank Level 7.

Sc rated Wate r Suppl, Temperature 4edundancy is provided for: Water level (in each tank) and system water flow.

Nitrogen pressure is sensed at toe cylinders and just dcwnstrean of the N2 pressure regulators.

The design of the system is not predicated on the availability of the pressurizer instrumentation and controls.

The ranges of instrumentation have not been provided.

The control aspects of s system are as follows:

The fi rst make-up ou,o will be cut in by a " Low" level signa:

'ry

+ > ci t o ?en w e SM IM surge tank; the second pump will be cut i n by a "L v L>"

-,a1 signal; and both oumos will be cut out by a 'High" level ;ignal.

The controls for the heaters have not been described.

L 4

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Je have required that the design include automatic isolation capability to preclude the addition of nitrogen into the Reactor Coolant System.

The licensee has responded by providing an automatic close signal to the motor coerated isolation valve initiated by low level in the tank nearest the Reactor Coolant Systen.

The isolation valve will be signaled to close wnen water in the tank is decleted to 300 gallons.

Tank c res su re taps will be utilized for level indication.

(One pressure tao in the tank and the other in the 6-inch piping upstream of the tank).

Because of the location of pressure taps and uncertainty of level control to actuate the isolation valve under transient conditions we requi re functional testing o' this contro' system that would require tank bl owdown prior to installa-tion of the pressure control and makeuo systen, i f t' ? test results turn out to be unsatisfactory other means of preventing N2 insurge into the RCS such as automatic isolation and venting of N2 supply will have to be considered.

We have further required that an alarn be provided to the operator when the di'ferential pressure of the Reactor Coolant System and the pressure control system exceeds a set value.

The licensee has complied and the alarm will be annunciated whenever the differential pressure is greater than 50 psi.

'de find the instrumentation and control a w cts c~ :- n sys en as de:crix d a;on to be accro;eola.

f ?

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i 3a.

4 Decav Heat RemovJ1 Svstem (Existino and Skid-Mounted)

Decay heat removal capability will be provided for the plant for its current operating modes.

To accomplish this, the following work and modifications will be undertaken:

1) upgrade the leak tightness of the existing decay heat removal system as required, 2) install an additional skid mounted train of decav heat removal equipment, and 3) consideration of a dedicated long tern decay heat removal and pr ima ry coolant cleanup system in a permanent structure.

A.

Uno ra d i ne Existine Decav Heat De~ovat Svstem Provisions will be made for conducting a preoperational test of each 1000 of the exis ting DHR sys tem.

Locations of system leakage will be identi fiei using television cameras installed at key locations.

Once leakage paths are identified, they will be corrected if possible, thereby oroviding as leak tight a system as is practical.

Leakage collection caoability will also be added to the system where feasible (i.e.,

collection of leakage around valves).

Instrurentation to detect JHR puro vibration will also be installed.

The availabili ty of the existing decay heat removal system will depend on the final leak tightness of the cr s ' G r ','g the c o t -.m i r a n i e n i n els p e sent

'r t % v r ie.3 y yy c. :

coolant.

2b3 041

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3asien of New Skid Mnunted Decav Heat Removal Svstem A third train for decay heat removal will be provided.

This will involve a tie into the existing decay heat removal system drop line d&~nstream of valve DH-V-3 located in the fuel handling building and ties I-to the two return lines to the cold leg-also located in the fuel "andline building (See ~icure 1).

< ew lines wi'! be run througH te aenetration room to an opening cut in the fuel bandling builcing sall and out to a skid located outside the building at grade elevation (304'-6").

This skid will contain a new decay heat removal heat exchanger and two

'ul! capacity pumos.

The discnarge line fo r the ies t exchancer will return through the opening in the 'uel handling b ilding to the return line tie-ins.

The tie into the decay heat remova drop line will be made bv welding an 1 inch weldolet a the pipe wi aa 'ull penet rati on weld, dye cenetration testing the weld, then cutting the hole in the oice using a clasma arc cutting crocess to minimize debris and finally ae'dino the new pipe to the weldolet.

A similar crocedure would be used for the tie-ins to the two return lines.

This procedure should minimice the time that the decay heat removal systam will be out of operation.

All va!ves wi!! Se e!actric notor ope 3'.ad arc v 11 hz c :o, sea:s with provision. for collecting leakage and directing it to th e existing rad-waste system.

Additional conne-t ions will be oroviJed in the new pipinc outside tne fuel handling building 'or 'uture u e in the installat:en o' a dedicated, long li#e, hardened structure which wili contai n heat 253 0<;;

t

. exchangers, pumps, deminerali.7ers and Iters for long term decav neat removal and cleanup of primary coolant water.

The seconda ry side of the new decav heat removal heat exchanger will be cooled by a new separate decav heat closed cooieng water system with i t s own pump, piping and valves.

(See Ficure 2).

This syste-in turn will be coofe by water # rom the nuclear services river water system.

New connections will be made to this system.

The design of this new decay heat removal system will be compatible with the current primary system decay heat levels and operating para-eters.

It will require that the pressurizer level be maintained half full at all times to insure acequate NPSH to the DHR pumps.

Alternativelv, the backup makeuo and pressure control system could s?rve to satis fy th. s requirement.

The new skid mounted DHR system and closed cooling system is comoletely separate and indecent from the existina JHR but is not designed against single active failure.

H owe v e r,

the existing CHR system would be available in this event.

The clan i s to ha ve the skid available but not make the connection to the lines penetrating the fuel aanol ing bui l d e r.g wa l l a t this time.

In the event this sytem would be reauired for incediate service, the skid and oicing would be shielded to provide all reasonably achievable radia t ion protecti on.

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The staf f nas reviewed the exnected system perforrance of this system and concludes that its heat removal capacity is su'ficient for decay heat removal from the IMl-2 reactor.

Mechanical Desion Considerations The skid rounted decay heat removal system consists of components such as heat exchangers, purps, tanks, val'<es and oicinc.

The apolicable design codes and standards used far the design o' these components are orovided in the table below.

Table 1 APPLICABLE DESIGN CODES OR STANDARDS Pumo and Heat Exchancer ASME Section 111 Cl. 2 Valves ASME Section !!! Cl.1 Picino M vtu e of Type 304 and 316 10" Sch. 40 stainless Steel sections.

ASTM Ma te r ia l Ce tification All welds fullv radiographed except for weld-o-let connections to existing OhR piping.

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. Dacicn Criteria Loads Considered - Normal operating conditions, - to include operating pressure, dead w -ight, pump vibration, thermal expansion, and maximum anticipated pressure surges.

Normal stress limits will be met for all piping and components including loacs 'com DBE.

Design Stress Limits Used:

Punos, Va l ve s, Heat Exchangers - as specified in ASME, Section ill for applicable Code Class.

Pioing - Stress limits per ANSI B-31.7.

Dasien in#or-ation-Sceci#ic #or Svstem Tie-in to E<istino DHR Pioina, Jeld-o-let Connection 7.einforcement area of fitting provides a 240 percent margin over the area c' the existing DHR piping it replaces Pipe succorts will be arranged so maximum stress levels at the weld-o-iet to DHR pipe interaction will be held to doout one third of the B-31.7 allowable stress limit for normal loads.

Weld-o-let to DHR existing pipe wolds will be made using a qualified procedure and by welders cualified on weld-o-let to pipe connection mockuos.

p

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s

. Secause of time constraints, the weld-o-let to pipe welds will not be rad i og raphed ; haaeve r, the root of the weld will be ground and dye penetrate inspected and the final surface will be dye penetrant inspected.

Additionally, the cesign is being qualified by hydrostatic pressure tests and bending moment tests which apply loads until the simulated existing DHR oine exceeds its vield streng!1 All welds and the cut into the existing DHR pipe will be performed using the plasma are method.

The plasma arc was chosen for the com-binat ion of small beat-af fected zone and mininun resulting slag which can be cleaned up with relative ease.

Miscellaneous Valves - Line valves and relicf valves will have leakage or discharge fluid piped to a drain tank in the auxiliary building.

Jacav Weat Closed Coolinc Water Sv-tem Comonnents ASME Fection lli CL 3 For all components Materials:

C.S. oiping 5.5. Pump, Valves, Heat Exchanger All :annections.velded except for piping to component interfaces which will be flanged.

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Evaluation

Conclusion:

We have concluded t h3 t the Licenece has speci'ied conoonents designed and Fabricated in accordance with acceotable industry codes or standards and will take into account t h e, loads associated, with startup testing, and planned system operation.

T"e use of corconents that are in cor#crrance with these :-iteria provides adecuate assurance that structural integrity of the Decay Heat Removal System will be maintained.

Structural Fuel Handlinc Buildinn Vall Penetration (Auw. 81do)

In order to provide the skid mounted decay heat renoval systen a penetration would have to be made through the west wall of the Fuel Handling Bldg., between column lines AC and AF and across column line A69, a t elevation 297'-0" (See B&R Drawing 2075).

This is approximately seven feet be l ow grade.

Excavation outside of tne structure will be done by pick and shovel to mininize a possibility of damacing any p.ioing or electrical conduit.

The outside wall at that location is rein #ccced concrete, S'0" thick.

The compressive strength of concrete in this wall is $000 psig.

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The coening is to be made 21" wide 5'1" high on the outside tapering to L'-1" at the inside face.

One reinforcing bar, 418 is to be cut at each face.

Figure 3 shows the f ront view and the cross-section at the opening.

Procidure At the t ime o' wri ting o# this report various methods of penetration are being studied.

One possibility is using the oxygen lance process. A trial pene t ra t i on is currently being performed by this method.

Drilling through the well is also considered.

The final method of the wall cenetration will be described upon pe oding resul ts of the test and con-sultation with the other experts in te field of construction materials.

Instrumentation for New Decav Heat Rr' oval System The following identifies the instrumentation to be provided for the third DHR System train. A trailer will be used to provide remote control room operation.

The follsaing instrumentation will be provided:

1.

CHR pumo suction pressure 2.

DHR oump suction temperature (OHR cooler inlet temocrature) 3 CHR Cooler inlet pressure 4

CHR Cooler discha rge pressure 5.

CHR Cooler discharge temocrature e.

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f.edundant measurements and display are provided fcr each oarameter and seoarated to be consistent with the A or S pumo train.

Independent power supplies are provided for each train of instrumentation.

No automatic control or interlock features are provided.

To the e x ten t oractical, the nuclear instrume":ation which is safety related has been purcne ed to Class iE requirerents.

The instrumentation selected was based on, to the extent available, the same manu f acture r,

model, and princiole as the instruments used in the existing DHR ;vstem.

The staff concludes that the instrumentation and controls to be orovided for the skid mounted decay heat system are appropriate for their intenced function.

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5.

Elect ical P owe r Svste, Modifications A.

General Descriotion The licensee has identified a number of modi ficdtions to the Unit 2 power system ir order to accommodate a loss of offsite power.

Backup power to the two existing safety re'ated load grouos (designated Red and Green) is supplied bv the two existing Class lE diesel generators.

.'lo loading cnanges have been proposed for trese busses and the staff posir ]n is that all new loads be powered from other busses leaving the original Class IE sy= ten intact.

The s ta f f has further stipulated that the loading of these Class IE busses be on a " manual only" basis due to the time available for ocerator action follcaing any certerbation and the corolex nature of trying to coordinate which loads are needed and which would be detrimental if actuated for each potential plant node of cooling.

Approved wri t ten emergency procedures a re recuired to cover the above contingencies.

'Je find the above described codi'ications to the existing Class IE syste, to be acceotable.

For all electrical loads that previousl y did not require loss-of-o s i t e powe r back-uo protection and all newiv added loads which now require this protection, the licensee is install ing two new diesel generators.

These diesel generators will be assigned to busses 2-3 (breaker position 3-3) and 2 4 (b reake r position L-13).

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These 1:ad grosos are now de s i gna t ed 'G rav a n d

'1.'h i t e ' ' r e s pe c t i ve l y.

The e diesels are self-contained skid-motnted outdes. :ni ts rated ncminally at 2500 kw ach.

The convent ion for bus assignment has been odd numbers for the A loop and even numbers for the B loop and this has been retained for the modi'ications. We have requirea that the

.e Iced grouas ae ~aintaired incecendent and t ct r c-i-tericcks shcl'

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• "at couples tha leac groJos in any *anner.

Each n w diesel will have its wn swit ngear, control battery, and fuel tank installed adjacent to it.

The switchgear will be connected (via underground cabling) to basses 2-3 and 2 4 (gray and white resoectivelv) through the existing circuit breake (previously vsed for condensate booster oJepe Co-o-2A cnd Co-p-23) with odifico relaying to accomodate the sawe r sources.

Loss of o site cower will be detected by new unde rvoi tage sche es wh ich wil l autorat ical.'. start :Se diesels.

The breakers at busses 2-3 ar.d 2 4 w

s r.o rna l ! y c! asco and the new diecel generator circuit breakers will be aut 4tically closed when the units have attained rated soeed and voltage. Motor loads are autocatically tripoed uoen Ivss r.' power as part of the existing design.

The staff requires that these busses be manually Icaded by anproved written orocedure.

Thes2 actions can be cer#craed from the control roon.

The basis 'or this re:uireae t is the time available for coa-ator action and t'e dierent loads required 'or the various potential

^1 ant conditio45.

The licensee's design is in con #ar ance vith this re- 'rament and is acceatable.

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Susses 2-3 and 2-4 have existing bus t rans fe r schemes that allow con-tiruity of power supply, shoul d one o' the two of #si te power c. rcui t s througn tre auxilia ry transformers be lost, by fast-transfer to the cther transformer.

This schene is to be left intact and the new undervoltage detection schemes will have a 10 second delay to accommodate the fast t rans fer if one circuit remains available.

We find the aoove modi'ications to be acceotable.

'ie w i l l require that tes t ing requi re. ment of the Gray and White diesel generators and thei r associa ted D.C. cont rol power sunplies (ba t t e r i e s )

be comparable to those that now apply to the Class IE diesel generators and batteries.

B.

Steam Generator A and 3 Short Te rn Solid Water Operation The 900 horsepower condensate cumps are the single largest loads that will receive diesel generator back-up power source orotection.

Tha circulating water punos rated at 2250 horsepower each have placed an additional restraint on the power systen.

This size load is too large for any of the four diesel gene ra tors now on site (red, green, gray and wnite).

The licensee has proposed to provide a 13.2 kv line from the Middletown Substation to accommodate the motor starting requirements of these large loads.

This line will have the capacity to start a second circulating water puno while supol ying cowe r to the #irst.

Cooling svste, requirements are 'ulfilled by one puno coeration.

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The new 13.2 kv of f site transmiss ion line will originate at the Middletown sunction Substation.

This substati on has a 230 kv and a IIS kv bus tied together by fou r auto-t rans forne rs.

This modification ut il i zes auto-t rans forme r no. 2 and connects to the 13.2 kv tertiary winding rated at 25 MVA.

The group operated air break switches on the 230 kv side will be ocened to isolate the :-ans'er ' om a di ect connection to the 230 kv system.

There is a spare 2 n kv line that terminates on the tower at the entrance to the Middletown switchyard on one end and terminates on the tower at the entrance to the Three Mile Island Switchyard on the 3ther end.

This l ine sha res a doubl e-ci rcui t line of transmission towers with an existing 230 kv line to the plant.

At the Middletown Junction Substation, a short line section from the existing caoacitor breaker on the 13.2 kv tertiary of the no. 2 transformer back to the spare 230 kv conductors must be constructed.

The caoacitor bank will be disconnected.

The neces sary addi t ional relaying will be provided for this cacacitor breaker to protect thJs new feed.

At the Three Mile Island Substation, a 13.2 kv underground cable supply from the spare 230 kv conductors will be run around the southern side of the natural draft coo l i ng t owe r.

This underground portion of the cable run protects the line from any of the other incoming lines falling on it.

Once around the cooling tower, the line goes overhead for one span and terminates on a 10 MVA 13.2/4.16 kv trans-former.

A space 10 MVA transforrer will also be in niace with manual switching s

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capability for energi:ation should the First one Fail.

These trans-

'ormers are located adjacent to the circulating water oumo house and the electrical switchgear bay.

The 4 kv portion o' the line ties into breaker cubicle 5-4 on bus 2-5 located within the switchgear bay.

This breaker was originally used to suoply power to circulating water cump CW-P-lE.

This breaker sill have nodi #ied relaving to acc-- odate the new c ow e r source.

We have requi red th i s al te rna t e powe r source schene to be

'manua l-on l y" by app roved w r i t t en procedure.

The bus must be cleared of all connections following loss of of fsite power prior to reenergization and subsequent loading of the circulating water pump.

The r e is a normally open bus tie between busses 2-5 and 2-6.

Closing this bus tie gives access to five of the original six our7s.

We will require that the 13.2 kv line be tested weekly by energizing busses 2-5 and 2-7 for a short time interval te assure continued

'unctional capability.

There are no phase angle dif ferences between the existing system ard this new line so that the connection may be made on a live bus followed by tripoing circuit breaker 23-52 to prevent tieing the 115 kv system to the 230 kv system through the plant distribution system.

We further require that circuit b reake rs T-56-2 and T-73-2 be veri fied open on a daily basis.

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1) the application is limited to the ci rculat ing wa te r pumps.

2) these pumps are only required for the time window necessary to comolete the modifications to steam generators A and 8 for long term core cooling, 3) the time required t3 have the 13.2 kv line operational is consistent with the requirement to orovide back-up powe as soon as practicable.

L) alternati"e propo als sucn as the use o# diesel generators were not considered feasible due to the sice requirements for starting 2250 horsepower loads and indoor service would require a new building that could not be built in time based upon existing schedules.

5) the nea l i ne ha-be :n isolated to the extent practicable from the 230 kv system and t erefore may survive a local 230 kv system disturbance.

6) given a total grid olackout, there are six combustion turbines in the cloce proximity of Three Mile Island.

These units are rated 23 MVA eacn and have black-start remote suoervisory control.

The system dispatcher in York has supervisory control of this system with the one exception that the system dispatcher in Lebanon would have to be consulted as to the cosition of certain intervening circuit breakers.

7) transmission line tower failure could not redner the 13.2 kv and the 230 kv systems inoperable.

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We 'ind the addition of the 13.2 kv transmission line as a backup power source to be acceotable given the operational needs and time restraints cresent at Three Mile Island Unit 2.

C.

Steam Generator A Modifications Thesc modi ficat ions provide a new 700 horsepower high pressure.iump for circulating -ater tnrough the secondary sice o' the A steam gererator.

This new loop is in cooled ov the Nuclear Services River Vater Svstem and the Nuclear Se r vices River Wate r Purges.

The new high pressure oumo will be powered from bus 2-3 (Gray), the Nuclear Services River Wa te r Pumos a re powered f rom the existing Class I E powe r system. We find the cower source allocations to be consistent with the separation of the A and 8 steam generator modifications, to caoable of supplying the oractical requirements of the system and to therefore be acceptable.

3.

Steam Generator 8 Modifications These modifications provide a new 700 horsepower high oressure cuma for circulating cooling water througn the secondary side of the 3 steam generator.

This new loop is in turn cooled by the Secondary Services Closed Cooling Water Systems which utilizes the Secondary Closed Cooling Water Punos.

The final cooling loop uses the Nuclear Services River Water Punos for circulation.

The new high oressure cump will be powered from bus 2 4 (wh i t e ), the secondary service closed cooling water pumos SC-P-13 and SC-P-1C are powered from LA0 motor control center (MCC) 2 418 wnich in turn connects to 431 volt bus 2-kl b

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_ 50 and through a 4160/480 voit to bus 2-4 This arrangement allows a back-up pump (one pump operation meets cooling system requirements) and is part of the white power system.

The Nuclea r Services River Water pumps are existing loads on the Class IE diesels.

We find the above power assignments to be consistent with the separation require-ents of keeping the B steam generator cooling svstem on the even numbered secarate busses. to be cacaole of supplying tne functional recuirements of the system and to therefore be acceotable.

E.

Skid-Mounted Decay Heat Removal Svstem Th i s new system will have 480 volt motor operated valves arranged in such a manner that there will be two sets of isolation valves on each of the three DHR lines that will be tapped.

These valves will be assigned power sources (Gray and White) in a manner that assures isolation cacability given a single power source failure.

The remaining 480 volt motor operated valves will be arranged on a "per loop' basis to allow selection of either of the two new 4160 volt h00 horsepower pumos.

These valves and the associated pumps will be powered from secarate busses (Gray and White) to assure systen function given a single power source failure.

Two new 480 volt motor control centers will also be provided. All elect ric cowered equip ent (valves, pumps, motor control centers and cabling) will be Class IE system quality.

Because of the low decay heat levels, sufficient time is available 'or manual operator action assuning loss of p ow e r or equipment.

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The secondary cooling loop for the DHR heat exchanger includes an additional 250 horsepower pumo connected to 430 volt bus 2 44 All associated motor coerated valves will receive normal power from the white power system.

3hould the whi te powe r sys ten fail, backup power ccn a provided from the gray system by closing the normally open aus tie creaker betseen buses 2 h4 and 2-34

'le recuire this tie cr?aker be racked-out at all times and oniv c!ased u;on the Failure of the white power system by approved wri tten ?rocedure.

This system will not be used concurrently with the steam generator cooling, odes describes above and therefore does not affect diesel generator capacity.

We find the electrical power aspects of this design as described above to be acceptable.

F.

Reactor Coolant Pressure Control System A,11 electrical equipment and instrumentation required to operate t'e svs tem a re powe red f rom the gray and white nower systems.

T"e charging ou os (A and 3) are rated 100 horsecower and are powered fro, 420 volt motor control cente rs 2-32A (gray) and 2-h2A (wh i t e) respectively.

The charging water storage tank heater is rated at 100 kw and will be powered f rom bus 2 h5.

There are a number of s.all loads associated with this systen that have not been ass igned powe r sources.

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The system will be autorated as soon as possible out this nav not occur before initial operation. A motor coerated isolation valve vill be orovided to autonatically close when the water level in the tank nearest the reactor coolant system is approxinately one third #ull.

This is to preclude the introcuction of nitrocen into the reactor

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6.

Qualitv Assurance Progra-for TM l 2 System Modifications The NRC Technical Review Group has reviewed and evaluated the Quality Assurance (;A) Program of GPU/ Meted.nd of thei r maj or subcontractors for the TMl-2 sys tem mod i f ica t ions.

These QA Programs recognize the uniqueness o' the TMl-2 clant condition and balance the schedule urgency for comotetion o' svstem modi'ications acainst the extent o#

a:oli:ation o' traditional (A procram p actices consistent with maintaining assurance that soecified system requirements are met.

The GPU/ Meted QA Program has been specifically tailored for the TMI-2 system modifications.

The Program w'll apply QA criteria of 10 CFR Aopendix 3 commensurate with the spec'fied system recuirements and will be comoa t ibl e wi th the Meted Ope ations QA Program oreviously accepted by NRC.

GPU/ Meted has estat ished a QA organization at the TMi-2 site soecifically responsible for the system modi'ica t ion QA activities.

This staff is exoerienced in all QA disciplines and associated :2chnical fields, including mechanical, electrical and civil engineering as w 11 as welding and non-desctructive examination.

The GPU/ Meted 2A Manager and QA engineers were brought in from the Forked River facility.

GPU/ Meted is the lead responsible QA organization for t he TMI-2 sys tem mod i f i ca t ion p rocram.

Thei r QA Program will orovice surveillance over the activities o' their subcontractors, i ncl ud ing w'es t i nghouse and Burns & Roe.

Jestingnouse has established a QA Program to control QA activities associated with cesign, pro-a 2 'J 3 pdub

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curement and vendor co,mponent fab r i ca t ion related to the skid mounted Sackup DHR system modification. Westinghouse has organized an experienced QA staff specifically responsible for this task.

Westinghouce will also provide system installation and pre-operational testing procedures and on site technical supervision which will be conducted subj ect t,

&* GPU/Me t Ed QA Program.

Surns & Roe is resporsicle for establishing designs for the other TMI-2 system modifications, including the designs for the Reactor Coolant P essure Control System, the A and 8 Steam Generator Cooling Svstem.

Surns and Roe is implementing design control QA practices which assure that appropriate quality standards are specified and included in cesign documents that provide for verifying or checking the adequacy of design and that will control design changes.

The GPU/ Meted QA Program will crovide QA surveillance for the follow-on activities for these systems, including orocurement, fabrication, installation and testing.

The NRC Regional O'fice of Inspection and Enforcement has available qualified QA staff experienced in mechanical, electrical and civil engineering disciplines and welding and non-des t ructive examination to provide surveillance of system modification activities at TMI 2 site and at equipment vendor facili t ies as necessa ry.

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'aluation of the QA nractices, controls, and organization of GPU/MetLc and their major subcontractors, we conclude that t he s e Q/s Programs will assure meeting the criteria of 10 CFR 50 Aopendic 3 co rnensurate wi th the THI-2 sys ten modi fica t ion requirements and are acceptable.

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