ML20132C938
| ML20132C938 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/31/1979 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20114E436 | List: |
| References | |
| NUDOCS 8504260474 | |
| Download: ML20132C938 (65) | |
Text
{{#Wiki_filter:. . . . . . ._ . _ - . ATTACHMENT 2 TMI-2 PLANT MODIF,ICATIONS ,
, t i FOR ACHIEVING COLD SHUTDOWN
- MAY 1979 i
I e i
*This report was prepared based on infor-mation 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.
i . 8504260474 790509 ,
; PDR ADOCK 05000320 '
P PDR i
TMI-2 PLAtlT MODIFICATI0'JS . FOR ACHI E'll'JG CCLD SHUT 009tl Table of Contents
- 1. Background
- 2. Steam Generator Modifications
- a. Steam Generator "A" Modification - Short/Long Term .
- 1) Design Concept ,
- 2) Modification
- 3) System Evaluation
- 5. Stean Genera tor "B" Modification - Short Tern /Long Term
- 1) Design Concept
- 2) Modification
- 3) System Evaluation
- c. Mechanical Design Evaluation (Steam Generator A/B)
- d. Structural Evaluation (Steam Generator A/B)
- e. Instrumentation and Control (Steam Generator A/B)
- f. Radiological Evaluation (Steam Generator A/B) 3 Reactor Coolant System Pressure Control
- a. System Evaluation
- b. Mechanical Design Evaluation
- c. Structural Evaluation
- d. Instrumentation and Control 4 Decay Heat Renova l
- a. Upgrade of Existing DHR Steam Leak Tightness
- b. Skid Mnunted DHR System
- 1) System Evaluation
- 2) Mechanical Design Evaluation
- 3) Structural Evaluation
- 4) Instrunentation and Control 5 Electrical Svstems Modi'ications
- a. General S. Soecif:c System Modifications
- 7. Qualitv Assurance
4
1. Background
In order to bring TMI-2 to a cold shutdown condition, modifica tions will be made to various plant systems in phases to be carried out over a period of the next few weeks. These modifications will permit a gradual transitten from the current plant operating mode to one which provides a stable long terra cooldown mode of operation. The first planned modifications will be made in conjunction with the transition from forced primary coolant circulation (by the reactor coolant pump) to natural circulation through the core. To accomplish natural circulation, the secondary side (shell) of the steam generators will be operated water solid. Water will enter through the main feedwater ring and exit through the main steam line. In order to provide for water solid operation, certain modifications to each steam genera tor secondary flow loop will be required. The design of modifications to steam generator "B" has accounted for possible contamination as a result of suspected tube leakage in the
. steam generator.
l e i
4 l - 4 l fo backuo pressure control capability of the primary syst'em during ! i natural circulation, a new pressure control and c:keup system will i ! be provided. This system is essential in the event of loss of pressurizer i heaters and level indication. Criteria and procedures for letdown and a j overpressure control of the primary system will be established prior to , going into this mode of operation. Because of suspected leakage in the existing olan't decay heat renoval ] l systen, a program will be conducted to identify and correct leaks to i, - i provide as leak tight a system as possible. Also, an additional skid h mounted decay heat removat train will be connected into the existing system as a backup. Connections will be provided to the new train for a possible addition of a dedicated decay heat removal and cleanup systen ! f located in its own permanent structure, , I , To facilitiate early completion of design and installation of these , system modifications, system functional capability following a seismic event has not been a design requirement. If a seismic event should occur and damage the modified systems, the seismic Category 1 TMI-2 Decay Heat l Renoval System and Reactor Coolant Makeup System could be used to remove i i core decay heat and control primary system pressure as necessary. l i i i More detailed descriptions of these modifications are included in i . i I . the following pages. ;
; I l
I i I r I h N f I
-a -- =- .,_
3-
- 2. Steam Generator Modi'icat nne
- a. Modi *ications to Steam Generator "A" for '/ater Solid coerations
- 1) Desien Conceot The short and long term concept for water solid operation o' Steam Generator A ha"e been consolida*ted. One concept vill now be ut!!I 'e i latse dLl' :e cireJ'a*ad Sv the nes l pump through the shell side or a new heat exchanger and into I
the secondary side of the steam generator A in a closed loop to remove heat f rom the steam generator. The tube side of the new heat exchanger will be cooled bv the existing Nuclear Se rvices River '. late r Sys tem (flSRWS) which suoplies water f rom the river and returns it to the mechanical draft cooling t owe r. Refer to Figure 1 for a schematic of this flow path. Provisions will be made for system pressure and expansion control by utilizing the existing 3rd stage feedwater heater i ~ shell and its nitrogen suoply as a pressurized surge tank. The design also includes provisions for sampling, demineralization and chemical addition cacability. For the initial phase of operation, all valves, will be manually operated and ] instrumentation will orovide local read out. The new loco of heat removal equipment has been designed to t ooerate at a nressure hic"er than the exoected reactor coolant syste9 pressure thus assuring in.-le3kage o# secordav system i I .
-4 liquid into the crimary system in the event of steam generator tube leakage.
- 2) Mo il f i ca t i on This scheme will involve Installation of a new high pressure train consisting of a pumo, heat cxchan$er, valves and piping located in the turbina building basement. The loco will be connected to the main steam turbine bypass li e between the connection to the main steam lines and the condenser, and to
] the main feedwater line between the feedwater pumo and 3rd stage feedwater heater FW-J-6A. Horizontal runs of piping will be supoorted for static loads and secured to suonorting structures i to prevent lateral motion. Vertical runs of piping will be secured to permanent structural members as required. Additional piping will be re julred for the surge tank (3rd stage feedwater heater FW-J-6A), chemical addition tank and denineralizer. In addition, the interconnections between the A and B feedwater heater trains will be broken and caoped off. Jumper pipes will be Installed between the existing Nuclear Services River Vater System (NSRWS) the existing Secondary
- Services River Water Svstem (SSRWS) to provide cooling water to the tube side of tne new heat exchanger. The safety classi'ication of the Nuclear Services River Water System will be maintained oy oroviding dou'le e Isolation valves.
l l i
. . _ . __ . . _ . _ . . . _ . _ _ _ . - . _ _ ~ _ _ _ ,
.g.
All piping connections will be welded. These modi fications are expected to be comoleted and the system ready for operation by the middle of May.
- 3) Systems Evaluations The system design as proposed will meet; the necessary secondary side requirements for decay heat removal with either forced or natural circulation through the reactor core. All required modi'ications will be made to accomplish 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 Vater 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 components. However, a single active failure within the system will not compromise natural circulation of the primary system In that the secondary cooling loop through steam generr. tor "B" will continue to operate (see staff evaluation of UKI-2 natural circulation performance). Tha flosrates predicted through each of the heat exchangers will orovide adequata cooling based on an assumed heat load of 30 < 10 6Stu/hr (RCP operating + 3 MWDH). Ooerating serforman-a and dasign parameters for the system are as
'ollows:
- . . , - - - - - e ---. , - . , , , , - . ---,
i j 1 1 Steam Generator A Modified System 1 ! Ooeratino Conditions Deslon l System I Location Pressure (oslo) Temo. (OF) Flow (com) Pressure (osto) 1 New Pump Discharge 670 100 3000-5000 800
!' 600 '
j New Punp Suction 500 IOC 3000-5000 ]
- (New Heat Exch.
Olsch. Shell Side) , j . New Puno Recire. 670 100 Pump min, flow 830 670 120 3000-5000 800 f New Heat Exchanger , i Suoply (Shell Side) i NSRW Supoly to New 100 85 6000 150 Heat Exchanger I (Tube Side) L I NSRW Return from New 10 95 6000 150 l l Heat Exchanger (Tube ] Side) The system flow arrangement has been selected to minimize l t fouling ef fects by maintaining Nuclear Services River Water on the tube side of the new heat exchanger. System operating r i temperature Indicated for NSR'J supply and return are design i values. ; i ) 4, b. Modification
- to Steam Generator "B" for Vater Solid Onerations 1 .
I I) Desion Conceot i l The short and long term concept for water solid operation of ,
\
Steam Generator S have been consolidated. One concept will be utlllzed, i i Water will be circulated by the new pump through the tube side of the i i new heat exchanger and into the secondary side of steam generator S l The secondary j In a closed 103, to remove heat Fr$n the steam generator. i i t i d _ _ . . , _ _ . _ _ . . _ . _ ._ _, . . _ . . - . , , . . _ _ . _ , . _ . , . _ _ _ _ - _ _ _ . , . _ . , _ _ . . . _ . _ . _ , . . - . . _ . . . _ . _ . . _ _ _ . .
- - . .= _ .- . - -. _ _ _ _ . _ _ _ - .
side (shell side) of this new heat exchanger will be cooled by the Secondary Services Closed Cooling Vater (SSCW) system which is, in turn, cooled by the Nuclear Services River Vater Svstem, by way of the ] f, mechanical draf t cooling tower. Refer to Figure 2 for a schematic of this flow path. Provisions will be made for system pressure and expansion
- control by utilizing the shell side of the existino 3rd stage feedwater heate u*t', i ts e <istiac *itrocen sucoly and thus will Sperate as a i ! oressurized surge tank. Tbc design also includes orovisions for j samplina, demineralizatinn, and chemical addition capability. For the initial phas of operation, all valves will be manually operated and instrumentation will provide local read out.
The first intermediate loop of this scheme (that portion of the secondary system which removes heat directly from the steam generator)
- has been designed to pernit normal operation at a pressure higher l than the reactor coolant system oressure, thus assuring inleakage of g
secondary system liquid into the primary system in the event of steam generator tube leakage. However, the expected mode of operation from 1 which to initiate natural circulation would involve higher primary side ) i i pressures.
- 2) Modification This scheme .Ill involve installation o' a new high pressure train consisting o' a pumn (LT3-P-1), heat e echange r (LTS-C-1 ), va lves Tbc 1,co will and cloing located in the turbine building basement.
be connected to the main steam line at a 10" drain not between the 4 i l
a main steam isolation valve and the stop valve, and to the main feedwater line between the feedwater pump and 3rd stage feedwater heater FW-J-68. Piping supports will be similar to those provided for Steam Generator A modifications. Additional piping will be requi red for the surge tank (3rd stage feedwater heater FW-J-68), chemical addition tank (LT3-T-2), and demineralizer. In addition, the inter- . connections between the A and 3 feedwater heater trains will be broken and capped 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 exchanger LTS-C-l . Jumoer pipes will be installed between the existing Nuclear Services River 'Jater System and the existing Secondary Services River Water j i System to cool the tube side of the Secondary Services Closed Cooling Water Sv< tem. The safety classification of the Nuclear Services River Water System will be maintained by providing double Isolation
. valves.
All piping connections will be welded. These modifications are expected to be completed and the system ready for operation by May 7, 1979. i j l
J 1 i 9 4 ! 3) Systems Evaluation i ! The system design as proposed will meet the necessary i i- secondary side requirements for decay heat removal with either forced , 4 J or natural circulation through the reactor core. All required modi-i '
! fications will be made to accomolish this purpose. This system is i
indeoendent and sacarate from steam generator "A" during all intended modes of operation with the exception that the Secondary Services J r I River Vater System will be shared by both loops of steam generator t I cooling, i The system will not be provided with redundant active components. l l However, a single active failure within the system will not comoromise , t .{ natural circulation of the primary system in that the secondary cooling , loop through steam generator "A" will continue to operate (see staff I evaluation concerning TMI-2 natural circulation performance). ! j i s The flowrates predicted through each of the heat exchangers will provide adequate cooling based on an assumed heat load of 30 x 10 6 Stu/hr , l , (RCP operating + 3 NWOH). Operating performance and design parameters for the system are as foll >s: ) . l \ ! h 1 I i , [ t h
, , . - , - - - - ,--,-.,- , .,.- - - , - ,--. - - - .,,-.,J---...-..,------, - - ,- - r.- - - ~ n e ,. n .,--.,,,-.,.
Steam Generator B Modified System Operatino Conditions Desion Svstem location Pressure (psic) Temp (OF) Flow (nom) Pressure (psla) New Pump Olscharge 670 100 3000-5000 800 New Pump Suction 500 100 3000-5000 600 . (New Heat Exch. Disch., Tube Side) New Pimp Rectre. 670
- 100 Pump min, flow 300 Nes Fest Exchanger Suppiv 670 120 3000-5000 000 (Tube Side) ,
SSCW Supply to New Heat 150 72 4000 Exchanger (Shell Side) SSCV Return from New 100 59 5000 Heat Exchanger (Tube Side) NSRV Return from SSCW 100 71 5000 Heat Exchanger (Tube Side) The system flow arrangement has been selected to minimize fouling effe:ts by maintaining Nuclear Services River Water on the tube side of the Secondary Services Closed Cooling Water heat exchanger. C. MecFanical Svstem Desion - Steam Geaerator "A" and "B" Modifications . All comoonents and supports of both nuclear class and non-nuclear will be designed or verlfled to have becq originally designed for the
- maximum loads that they could be exoosed to during testing startup, ar.d expected operation of the system. I.e., pressure, temocrature dea dwe l gh t , ouma vibration, etc. TFe component design structural Inf3rmat ion is listad in Table 1
F %',*'T e.w tao L . n,p g,ug ' W J ' Q W l i ' e B. I ggg , %
- d. 5) l I
k [g 5 _ [{q'<c 5 E l 3 ww gE
~
~~ ~ @s o l ew o (h a'
) l l .f.
u E & E' E i 6 3h __
, M, vu u r.O $.;
\
% e M C+ A l I
l $f s %, I i
?
i u j
- w. '
f " i 2 g g l h W+1
~ '~~ ~ ~ ' ~
j
' . . . _ .._._ . . , i _
y ;_ . ), ___ ._ . , __ __.______ __. WFIA-dl . Vl+4 U vdA. .. M .
.rL W
_ . . l .- .
- po-mJ l
. . .p. ' . _ . .
. u
+
- - . . TJ
_par ; __ . i T.I__.. _ - . - -l__ ._ ~ . . __ gd t -~ ~ ~ Y-I
\
,i 58 2
- 1 n -
s . 1 J e a [ m-vira i I G/-{lmM v74 h TE
. . . s 5
i i i
. _ . - _ _ . .. . _ _ . _ _ . . .__p__..._.-___... .
4 4
=
. .. . .- . -..-..._- - + . . . . . . y -- ._. . . .
1 _. . _ _ _._ 9
.( ,g. ,
I _ .- ___ ___ g%1 ) j ._ . _ _ . l . , _ _ - I... g . . -. ._ .. _-..._.L _ _ . , _ _ - . - . .
- r . . _ . _ _- _
r_ ,
- _. ._. - u .. . _ - _ _ .
3 - u, + -.p --+ ._ s g_ , ,r -- J. u G}-][ 9-]Q [ -- -M - K_- . ..~j,T] g' _- E. I -- i g _ j _. i .. o t c l
, _ __ . y g 7,,m R 6e
- I I_.
.g __ . _ _ . g _ _ - . . . ... _ g- . _ _
- .:n ._L___,..._.. ,_ g ._..
.y C . .- . . _ _ t ._ . L_- : -. ._
[. . - - w _ & 1. .
._ ._ q- . . - . _ . _ ... _ __. ._
.g w .g . _ _. d . . . _ . - ._ . . l ._ . .
qL . L_ y . l _ . ._ _ ] o s l
.__-_._.I.._..q[{_
b) e p - - - y 23; l cg
- s g --
I
,. . g . . ..
.-- p g g -
.. a 2 p'4jdE 4 e-Ec y @
u 3 5 O 1 3'S8 psy v .
' I L-( f, y
i 3 2' T YI? I g ej k l ) ,
$b -
>I s n_
I b
- c' I hhk.
l
#(
yf-oi smoos adv.aC MfM/HUMl % 4 l _jf3 r ',
. s 1 m ) _l k
_ _ . . -m - . - - , r_-.-- ,.,--,------,---,.-% - - ~ ~ - - - , - , . ,w-.
- - - ~ _ _ _ _
h e;laaV (a < 1 y" M
- I D. Q'- %. l ' r-..- - t 3
Im 2 .,ri @.U. 1 f a . rJ Tmem. ~ i u. Twe W qa T 1 . I n r _ALI rN k2 s
)
' {F"-
- md 1 (4 4
)lw.1g ga 5 Oi t
= -
u ,e i 5 I ~ ,irs-vns EE ~ e jel3 3h I I - *- h- 13 l l- 3=a w s
$C dia d
1 ; ; u l r- - -+ 0 r- y f, Ag 5 ~-*- ga ,, r-J i-I- 92 r3'-y , w o 2 E O=!._ l L..". m k- *x =. A -
~
- = = = = =
c-- + ms ; e I i i IA u h, i 4 *
"e 538 4 Sk
?
l i Ti
'!' 2D IIt
- j *a' -Is$ h i
b d$ a
* - i 71 eI as w
% r1c -
Je? y d w
- y O> ~4h y [
J t g${ ~ ~ ~l L~ %.t4 T , s e L] rl ' by 8
--0 -* . M 7 r %, -
v t-*'*?i he l.,J et++ - - ;d t- Ean L -e i w -
- _3.n t* 2 ed s $, 2 a
T r 7 M'5 g .sgo l U> 2Ne
$yy 4t sw q E e a, , e,,, *
$ ?- g I O _1 }
{"
* -J L- - g" ["' -.- M- - ' J s, y =, .i , , , >
e2 a c 1 2 pc L^ s?v
-a r 11 b g -
t4 m
%2 3 P ', --
' a I s* s p~ 'a v4
. %; > g">g o s r. .
Ta h O # ,,v T ,j, M . J g ! c2 $4) -
-? w 0*= T 3 -- E q b wwl y-gr ...s
} 8C 2ylt qv
, m -ol,c,y>2 *m - -_ ._
(, b WsA-is ; O
- h. h, e m e Qy W
M i o e 4 N l g w U c. E wc. 4 2 y I
- w W I NE~ o G y " \
3 2 V u C ' r pec I 02 q T k'- re e d. ' i
, y c koG.$ T i g
- t- W 5 I C & _I 3
.$. - I r
. -- , ~3 \,A l - , 7g I
s k 'v'7 l
~ I - l
- II - } ,
5 i '
! the structural 4 A specific concern that we have addressed in our review it 7 \
adequacy of that portion of the Main Steam piping system which as in-corporated into the OTSG cooling system will contain solid water in l Components ? lieu of the pressurized steam for which it was designed. j Special in the system will not experience any significant dynamic loads. - /,
~
i precautions will be taken during initial filling and startup of the Loads that will system to minimize the potential for water hammer. be experienced include pressure, deadweight of water, and thermal 1 expansion. Since the system will be operated at a maximum pressure i of about half the design pressure of the piping and its maximum i operating temperatures will be considerably lower than the design 1 1 temperature of the main steam pipingi stresses resulting from these ( loads will be minimal. i the existing A'ter assembly and prior to initial operation of the plant, Thus 1 ) piping was hydrostatic tested and at that time was water filled. the oiping and its supports have been demonstrated to be adequate for i . the weight of the water. I , in order to minimize piping deflection, the licensee ' f Ve concur with I i has specified that selected spring hangers be pinned. i this requirement. 4 t i i P I i
- k i
i'
l
- 12 - ,
i Design Bases Lnads I It should be noted that all ASME Section 111 CL. 2 components used in i ! the cooling system were designed for seismic Category I service. How-j ever all of these components, both those that are part of the original i . TMl-2 Main Steam and Feedwater piping system and those obtained from , other nuclear sites to be incorporated ir.to the OTSG cooling system, j are being utilized in a system with different response characteristics I from that for which they were initially designed or are operating with l i a fluid media dif'erent from that for which they were seismically quell-4 I i fled, i.e. , some components designed for operation on steam during a
- l seismic event as opposed to water filled as in the present system. Thus because of these differences from the original seismic design requirements, i
i which can affect seismic response, these components should not ce considered seismically quallfled as installed in the proposed cooling system, solely on the basis of their original qualifications. Additional work would ; be required 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 evaluation of this system is planned. Evaluation Conclusinn i I We have concluded that the Licensee has specified components designed i and fabricated in accordance with acceptable industry codes or stan- ) i dards and will take into acccont the loads associated with startup, T l testing, and the planned system eneration. i i l i i t 4
. . - , ,-%--.. , - +. - - - _ _ - . , ,.-v. ,,,-e <,m., , , ,. -r..--,_._,,,.,,,_._.,,-w. .__.--,-,,%.-
. .y- ._y,,,-.r-r.-ryr.-
13 - l The use of components that are in conformance with these criteria provides adequate assurance that structural integrity of the OTSG "A" and "B" cooling system will be maintained.
+
l e-O t l i
I i .
- Table 1 Component Desion Structural infnrmation 4 .
Pumos A & B Svstem ASME Section ill CL. 2 i ) Derign T a;;r.aure - 350 0 F . Design Pressure - 700 psig. . Heat Exchancers j i A Svstem i l ASME Section Vill ; i Design Pressure: 150 psig (tube side) 4 600 psig (shell side) j 1 i B Svste, 1 ASME Section Ill CL. 2 ) Design Temocrature: 3500F (Both Shell and Tube Sides) Design Pressure: 675 psig ( ube side) ] 200 psig (shell side) , 3rd Stace Feedwater Heater (shell side as surce tank) 5 l ASME Section Vill l , Design Pressure: 1000 psig i 1 j Steam Generator Secondary Side - ASME Section ill CL. 2 l . Plnino. Valves and Mis. Tanks (minimum requirements)
, Ploing - ANSI B31.1 Misc. Tanks - ASME Section Vill Div. 1 l Valves - ANSI B16.5 and 316.34 1
i Frem Faedwate r check va lvas out s ide con t a i nment t3 the main steam ! Isciat inn valves i-l ASME Section !Il CL.2 Suc, orts ! ANSI 331.1 or as described in text. l
. 1 j
I s i
,-y . ~.. ..,, ,,,-,,mw -
,*,rw. -%---.,- um. ,--.,,v..-n, ..w.+%9%g ec-revy,,,,,v,my..m.-wwm-y..,r-,_ww- .. --yr~ +- -
15 -
- d. Structural Desien - Steam Generator "A" and "B" Mndi ficat ions Modification of Steam Generator "A" and Steam Generator "B" involves installation of one pump and one heat . exchanger for each steam generator.
The new equipment will be installed in the north-west end of turbine building at elevation 281'-8", between column lines TJ and TK in north-south direction a n d Th l a nd Thh i n e*a s t -wos t direction. TFe encloseJ Tinure i shows the general area of locatinn o' the aboie a equiona,t. The details of the eculpment is as follows: The heat exchanger for steam generator "3" weighs 36 kips wet and is succorted on two saddles , 2 8 -0" by 4'-0" each. The puro for this steam generator weighs 13.7 kips. The heat exchancer for steam generator "A" weighs 135.9 kips wet. It is acoroximately 33 ft. long and is supported nn two saddles 0 in, wide and 4'-9" long located 13'-C" apart. Examination of the existing structural drawings of the turbine building area, where this equipment is to be installed, reveals that the base slab is four 'eet thick, with #11 rebars at 12 in, scacing, each way, too and botton. Too of the structural concrete base rat is at the elv. 271'-0". The base -at is covered with a 3'0" layer of lean concrete 'ill plus appro ciratelv ' in, wearine sur' ace, Sringing the too elev. to 2 ? I'-?". Ass. ming L9 cegree distrii2ti_a n' the Icad, a-d sa : e basis o' t"e atiewable concrete :earino n-esrJre, the
I lI: N @ll j f -o!
'[S ,. ' ';p gj E 4l ,
g
-Q' -
s k. c. ts _. ' 's .R= =T t . CO M.'t l' A %l t t i qgs 2C 9 %@YG 4'0 r L h T i. kg,n C D q) % 3; e E &h
- s. . i = u 4
q . IN ^ y m ng j gC ((D q s a' es
'% L s s*
n 6 0.n2 o ' -
"x- s b
U h NN 0N w -$ u @ % 2 y ! O .C o , N y .t U N b
- l nM -
\ ,
n : .\ - p Ni l;,i!a . sr 7 _ . . v._ . . cli . 3._ _ _ g.. . . ._. : _ 1
; __ g L .__.c.__. . p. ._ ~;
.,___ _ v -- !
Ic' r w +;-
-- 59, ) L t -
1 :: - i i I i .1 . @.sm " ii- sjl^ " ' 6,.lrl.];ilh;
'. is -
L j!.fi,ij x i l.1
, i 6 *"ww&
) - .
j . i i I I n l.ir- 46 t ,e. e i
;i
- n. t as I ..
I ^\v 3 ) l i l llI --i$ ' ',Ih '- - II :l!g bil 0, h 3 I I P-!! M;; s . sn- y: - 1T1.;.._.. .;
,I y
--g" . c, . ,, i , fr m 11
! E 'h c i i h' I ':l-d'.! l f.i ! 9-c I
mL l @ f n~blr b x,.
-- I i
n
.. i.
!I ! it
- 9 i -
i. P o., p _:. . . j y + 2 I; 8a. . - . g, lj;et ;; II ,b.,15;: oy I' .: E i1 a . . . ; .j o j lIjl I t il G UI l ! lF q illill 3 '_,.,g;. j:jp; e
"E.
i!' o h !!i H I N .pt
- 4 T_q _
I y__.Lqi._ke=FhN,i.0U;rl i h:L.w
, c.,-
h:,+t a c t-u h;v l
-l 3 ..
l[s r. [ht - l- 3q,,g I
}
j
' t 'i
- d :' r .eq lg I t, i ul .m, h H !j% ! '! 4.c,,.3:.,3
-pI.: p., .e, q ! g ,_.J -.d w /,,
1 . -l .a ]
=
p , - - E g, r j g
!, a d. -
i. i
*]a:;, F.~.d l .:j i- r-e d. '3I m ~ da G
v T: E-" *N.p:]lEiI u .
- 1; ij ep l p.w) g> ,.J i ,
. M.2...ME.Ip[+ p (-
tkj $ H ys !
- b., ' - {-
.$_ T ,, see 3
. , p .' a . --e d -
M
... , E. e g.p.h_l ; > _t mt^n
._, e -h;u,c
- , % .__ ,_. . px y7; p;_
l, g , , a. .__ _%s._.c 1 _ .1. . . ,;a
- . s.t.
=m
%. \L.. w.,
. _ .,__wi..' c:. -_
}. r--~ , I "-.,%u v; 7' .
r,1.n __ l s -
~~- ,
1; 1- d' y* **MI i
, -l ;j - --).
s . i E.a.
. .s . .l,e.
,~ 7. , '
d-- i.. N.
- w. -
u j :. h W /-fv. , -3.g' . .-oII s" 3 N.:s. u h . I. ;* .,;-
---[7]---- 'N g'}' / - - - r- ::- .
j
. , t . . ;_ . . >w i,'_..___.-;__,\.._._,'
h j ' /'., h [ a9 d N . l l e N m;t, h3., y j!ITIgo,g c..-4' -* s% .
,l l':.l..; -
.__ E/ 'Ii x n _ q' ie m v .: . n
, i x o: __.
w.a w n, - =r. 1i-: *- u,=p :.r . e_ j s_ u x , q i , 'r., =
.v - .e a _' . c
'j .
- a i
-,,y- - -., % {" , ' ' _
i ,' D-.-_~ r. '
, .jM .
i i - w. E.7.2
.~~ - ]' '.
. R,yL s _-- ...,w_,,n..__._~
)i i
]i
' I I ,
==
i
, . i l-G '
l Jl !
- i i r
k
- . . ~ . - . - . _ - . . . . - . . . . . - - - - __ . -- . - - _ - . _ . - - . . .
i 1 I i
- 1 ,
I licensee estimated that the maximum load that can be placed over one square foot of the area is 90 kips. Since the heaviest piece of i equipment to be installed in this area is the heat exchanger, 135.8 l klos, which when distributed over the area of the supporting saddles ; i (2 x h.67' x .67' = 6.1 sq. ft) delivers the head in the floor of ; i 22 kios, the IIcs.,=ce concluded that the stah is cenable to supoort i the load. The analysis was performed for the dead and live loeds entv, under static conditions. We have concluded that the licensee has performed his analysis in accordance with the methods and procedures which are soecified by the I appropriate codes and standards. The use of these ce thods provide i a reasonable assurance that the structural elenents affected by this l 5 ' nodification will perform their intended function. 1 , i e. Instru-entatinn and Control (OTSG "A" and "B" Cool ino) i 1' The following new instrumentation (listed below) is being provided on the OTSG "A" and "B" Long Term Cooling Systems, i Radiation Monitor (* 'B' only) ~ 1 l I
- 2. Loop flow Indication *
] ! 3. Steam Generator Heat exchanger, Shell Side Temperature - T;n i ' { . (*A on l y) , T out i L. Heat exchanger, Tube Side Tercerature - TI,, Tout
! c. Puno Pressure - S;ction and 31scharge-i 4
- 6. Loop Temocratu-en
?. Loop Pressure
- Redundant sets of instrumentation are marked with an asterisk.
l .
Control room alarms will be orovided 'or radiation as a result of high S/G out-leakage. For initial installation and system operation, only local indication will be available mounted near the sensors in the turbine building. Sinilarly, all controls will be local. S/G leakage iato the coolir.g locos is to be sensed by victoreen model 355 area monitors strapped to the piping just down stream of the tie-in to the main steam line. Loop flow rate is to be sensed by 2 (two) barton model 200 differential pressure mechanical indicators across a single permutit orifice plate. All new temperature sensors / indicators are speci Fied to be ashcrof t (5 - inch code 53 EI) bi-metal, liquid-filled l thermometers. Loop pressures are to be sensed by ashcroft 1279 (bourdon system) nechanical pressure gauges. Table 1 provides additional instru-entation characteristics. The system design 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 control room have not been developed at this! time.
I
.s j OTSG "B" C00LilJG LOOP - TABLE 1
- VAR l Alj.I E SENS0k TYPd LOCAT10tl S/G "B" t. cal. age Strap-on gamma Detectors I monitor nn existing main steam turbine (primary to secondary) Victorcen bypass header ("A" loop only)
~ t 2 monitnrs on new pipe just downstream of existing main steam heacer ("B" loop only) Lono flow Tcta! Finw Orifice Plate-Permutit 2 indicators in new pipe Just downstream of ex i s t i nq ma i n s t ear. I-ader ("A" loop) 2 indicators in new pipe just upstrea,n 4 of existing cedwater header ("U" l oop) ' Bimetallic in aev piping juct off mainsteam header Thermomete r-Ashcrof t x i < 23- .Seng-r ,
/ .. Side, Tempe ra t u re 8imetal1ic At heat exchanger Thernometer-Ashcroft
~;n, Tou t Bimetallic At l'ea t exchanger
!..ae Side, Temperature Thermometer-Ashcroft Iin. T out l Pump Pressure a
Soctinn and Discharge Pressure Gauges-Ashcroft At discharge and suctler. of pump L.>op Taapera t ure Bimetallic Downstream of pump recirc. Thermometers-Ashcroft L.wp Pre. sore Pressure Gauge-Ashcroft Upstream of heat exchanry i i 2
- - . _ _ - - - - _ _ _ - . _-- --__ - w - --
- 19 -
- f. Radioloolcal Evaluation (Steam Generator A/B)
The radiological considerations which are integral to the water solid steam generator secondary side cooling method will assure that radioactive effluents from contaminated systens will be controlled and
- minimized. Additional precu3tions will also be made to minimize occupa-tional radiation exposures to operating oersonnel.
The secondary coolant presently la the "B" steam generator is contaminated due to the initial primary to secondary leakage which occurred on March 28, 1979. The measured radioactivity concentration and curie content is estimated in Table 1. Under normal conditions the secondary coolant pressure will be maintained at a value greater than the primary system pressure such that if steam generator leakage flow oaths are available, the highly contaminated primary coolant will not enter the secondary coolant. However, it is expected that transients of short duration may occur such that a reverse pressure gradient could
- Introduce additional radioactivity into the secondary coolant. To alert the system operators of such a condition, Indicators and alarms for pressure and radioactivity in the secondary coolant have been provided.
These indicators will alert the operator of an adverse condition so that corrective action can be taken prior to significant additional contamination of the secondary coolant. i l t
20 - The steam generator secondary coolant system and secondary services l closed cooling water system will be periodically sanpled 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., increase in the steam generator closed coeling loop surge tank levels. Leakage may also occur at various mechanical connections in the secondary cooling systen. To the extent practical, incations where leakage is likely, e.g. , valves with a known leakage history will be evaluated and a progran to minimize leakage implemented. Even with a leak minIntration progran leakage of contaminated liquids to the floor drain system may still occur. This leakage should flow into the floor drain system and be collected in the turbine building sump, the turbine j building sump pumps will be operated in a manual node with an analysis of the sump water radioactivity content being made orlor to sump pump operation, if the radioactivity content Is low, the water can be discharged through its normal flow path to the river, l' the radioactivity content Is high, such that discharge would exceed technical specification Ilmits, l the sump water will be pumped to an appropriate redwaste system for treatment. Adequate free volume in the radwaste system will oe provided l , for such contingencies.
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 flow of contaminated water. Gaseous ef fluents f rom this system should be negligible. The noble gas inventory in the "B" steam generator is negligible because the steam generator was vented. Airborne radioicdine releases should also be negligible because the secondary cooling system is not vented (a nitrogen blanketed surge tank) and the low secondary cooling system temperature (1000F) resul ts in a low ai r/watee partition factor which reduces the volatility of the radiolodine. The licensee is oroviding a demineralizer system to maintain secondary water chemistry and to reduce radloactive contamination. Details on this system are not available at this time. The denineralizers will be shielded and designed to pernit resin bed disposal as radioactive waste. The shielding and spent resin handling system will also be designcd to minimize occupational radiation exposures, e.g., the use of disposable demineralizers. Following the THi-2 Incident there has been no indication of primary coolant leakage past the "A" s team gene rator tubes. Since the primary
~
to se:endary system pressure dif'erential will result in in-leakage only, precautions similar to tho".e evaluated above 'or the 'B" steam generator modification are not required. P
Table 1 "B" Steam Generator Radioactivity Content (Activities based on 4/19/79 sample reported by B&W) May 9, 1979 April 19, 1979 _ nautoussavacy Qj gij (Aj Half- Radioactivity System System concentration ~ Life concentration Inventory (uCi) Inventory (uci) (Ci) gm (Ci) gm 9,9 8.05d .93 50.8 8.9 x 10-2 I-131 ~
~3 .59 5.3 x 10 0.59 Cs-134 2.ly 9 x 10
~3 .54 1.6 x 10" 0.18 Cs-136 13d 8.x x 10
-2
-2 2.2 2.0 x 10 2.2 Cs-137 30y 3.4 x 10 7 '
cc (1) S.G. water level = 358 inches = 6.5 x 10 cc (2) S.G. full at 625.5 inches = 1.1 x 10 (3) Assumes decay from 4-19 (21 days) plus dilution from filling the steam generator.
m._. _ _ _ _ _ _ _ _ __ _ _ . . _ _ _ _ _ . ._ ___ . _ . _ _ _ . _ _ _ 4 i I
- 2 "l -
j I 3.0 STANDSY REACTOR COOLANT PRESSURE CONTROL SYSTEM r A. Svstem Evaluation I
- 1) Description I,
A standby reactor coolant pressure control and makeup system has ; j been proposed by the licensee. This system would serve as a backup to the J } C'ICS and maintain reactor coolant system pressure with the pressurizer 'Illed a , solid with vater'. Primary coolant system pressure will be maintained f { 1 i even with the loss of pressurizer instrumentation and inoperative i pressurizer heaters. Also, the pressure control system will be designed
- to provide adequate NPSH to the reactor coolant pumos 1f they are caeded.
i 1 The stanby reactor coolant pressure control system will consists of i j passive components (a series of water storage tanks and a surgetank with nitrogen blanket for pressure control) and active components f l (charging pumps). The system will control reactor coolant pressure i I over the range of 100 psig to 753 psig. t i The passive reactor coolant pressure system which would be operated initially 4 ] 1 l with local control, Additional Instrumentation and remote control will i be incorporated to permit ahtomatic 6peration of the system. The active i i pressure control portion would resupply water to the surge tanks with added capability or providing additional makeup water directly to the ! RCS If needed. , t . 1 i I , I l t I l- ! l :
. . _ - . _ _ _ _ _ . - . , _ - _ - .~ _. ,_-. _ . _ _ _ _ .._. _ . _ _ ,. _ _ _ _
- 2) Modifications The standby 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), nitrogen bottles, two 49 gpm positive displacement oumps, a This system wit!
degassed borated water sucoly tank, valves, and oiping. be connected between the di-cFarge side of the high pressure makeup system downstream of valve NU-V-144C and upstream of valve MU-V!6C. All the components will be placed in the fuel handling building. The above modification will establish a flowpath 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 barated water used in the pressure control system will be provided by the present chemical addition system. Connections will be provided to accommodate the addition of boric acid, H 2, demineralized water and hydrazine, LIOH and NaOH. The degassed water tank will be replenished via piping 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 I depicts the proposed pressure control s/ stem and interface connections to the existing systems.
I eQ SPc-7-1 V 40MV ' goo sas. rn Suna E~ PT- ->NN RL FsLL , E
" "E 8
;,o sec_y.4 3 ^
T Y e A p .y$ 7+Aurn f 1INE $ PA RV ! t or uso at
- l T w u v44,r ggg
- - -@ y
#2 boric RCID i
g [ c en rcuixa
- zy n. G = arm fryp) ~g*'AMK 4- b T -I Y - -C & 'A u p-vA44C N I
y p(, {Anu-Vle; E 7 ]L- ,. _ g y
-~ ~ ~ ' ' ' ~ ~
f
-~~
Hr1 4~ _. n
*** ~'
+ % + -{s 't-~~ ,, ",'llS c
~T~ 5 g Aup-T'4) Mi lg40* P-/S f , 'JG---
i SPC-7-4 MU-Y2LS s J L
.l. ,
y Dr&* n'
- T #'M A f 1 bcMnTre --
1
> warta l
l g : ; a u (,,,,g M f Bo naricswerc l ) *Txir,e.
- . n a a.- .n gn LPS p, siE R 1 w --4 k p{. _
--kC E 4a o**
$?C.P-1A
,s M
l A g girg I ft'li s:ar-:. TING - m.
~
g- W E:06 Elb6 - n 4D ju t '9 f - - - zu s riva s ys I ; * * ' *
- nt w sy3 f~fG irjQ[ $, $ 78 AfD8}f )tEACTOR CeCL ANT PRESSR E CwTRet QS TM
I i I
- 3) Evaluatinn If required, the passive portion of the standby pressure control and makeup 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 translent e
events that can cause considerable shrinkage in the RCS. Because of the 'Inite in eentory (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 4 gpm for 8 hours. Sufficient makeup volume requirement can be met by the proposed RCS pressure control systen for moderate system perturbations 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 following a hot leg temperature rise of 50 F. e For this event the licensee has shown a total volume change of 1900 gallons whi;h can be made up by the proposed system. H oweve r , the pressure control system Is not designed for makeup requirements of more severe translents such as a sudden em,plete loss of natural
- circulation for a period of 2 hours, followed by an RC pump start.
Procedures which permit reactor coolant pumo restart following loss or natural circulation would require tF; avallability of other makeup systems such as the HPt in addition to the orenosed pressure control svstem, h.
The reactor coolant pressure control and makeup system wlll not provide letdown caeability of reactor coolant caused by overpressurization events. Reactor coolant fluid expansion will be relieved by one or a combination of the following current components and systems: (1) normal letdown line (through the letdown coolers); (2) maintaining letdown with concurrent i termination of makeuo/ seat injection; (3) continued reactor coolant
- aura seal ret arn Fl ow; (4) opening o' pressurizer vent valve; crd (5) li' ting of the pressurizer relief valve.
Piping integrity of the reactor coolant pressure control and makeus system has been examined for postulated overpressurization events such as inadvertent startup o' an HPl pumo. This system will have a design pressure of 1000 psig except for the section f rom the HPl makeup line back through the second Isolation check valve which will have a design pressure o' 1500 psig. Oserpressurization protection of the latter oloing section when the HPl/makeun pump is started w:ll be orovided by installation o' a reller val e about the HPl/ makeup oumo set at 1000 psig and a reller Flowrate of 525 gpm. Check valves located d>dnstream of the HPl/ makeup pumps that are inside the Reactor Dallding will also provide
- protection to this system from Inadvertent overpressurization of the RCS due to any other causes. The criteria and procedures for letdown and overpressurs control o' the RCS will be established peloe to the operation af this syste .
i
_ _ _ _ . _ _ _ _ _ - ._ .- _. .. _ _ _ ~ - _ -_- - I . 27 The proposed reactor coolant pressure control and makeup system may l not meet makeup demand foll>< lng a depressurtration event of the primary system such as inadvertent opening of the pressurizer PORV (wlth downstream Isolation valve open). l( not Isolated in time, the system would be ! drained resulting in entrainment of nitrogen into the reactor coolant svst19 Ta 0 event su:h an sc:;erance the system will aut7eatically is71 ate on low level In the water tank. In the event of loss of o## site power I this valve will fall 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'fsite power. Also, an alarm will be annunciated locally and in the control room when the Isolation l
valve is not fully open. We have examined for single failures that con disable the pressure control and makeup system. The discharge valve SPC.V9 Is a single failure point, t h 7w t v e r It wlll be a manually operated ball valve p1sitioned to an open i position, and tSen locked In that position. For simplicity of design
! and Installation, we have not required redundant valves to meet single criterlon to insure system Isolation capability. A redundant charging pump will be available to fill the water tanks in response to tank . level reduction.
i { l l t
- - _ . . . . - __ - - . _ - . - . - . - - - _ - - _ _ - _ - ~ __ -.
.28.
4 1 1 i Requirements for preoperational flow test of the 9C$ pressure control
! system will be determined pending a review of the margin suggested by l
analysis for peak makeup flow demand and what the system is designed l l to provide. 4 Ne conclude that the proposed system will maintain reactor coolant - pressure control for normal water solid conditions and provide sufficient i makeuo water for a wide range of espected transient events that would i l cause shrinkage in the RCS. Emergency procedures will be u ed for t l additional makeup capability that may be required to mitigate more severe and less probable transient events. 1 l , j 8. Machanteet Deslan ] I f) Deserfetten The Reactor Coolant Pressure Control System is madeup of several f
; water supply tanks, positive displacement pumps, valves, and piping.
j TFe applicable deslan codes and standards used for the design of these ( 1 ! components are provided in the Table 1. i i , i 1 i i k I i l t r i l i ! i
l l l Table 1 Aopticable Deslan Codes or Standards l Water Sunply Tank (Passive Sys tem) ASME Section Ill CL 1 - 5.5. Charginq P3mps ASME Section ill CL.2 Pleing - ANSI B31.1 (minimum requirement) Decassified Vater Supolv Tank ASME Section Vill Div. I Corsonent Supports ANSI B31.1 1 ; 4 I l I L
. t i
l i l i I
_ . . _ _. . _- - .- _. - _= i 30 i
- 2) Design Loads Pressure, pump vibration, component and fluid deadweight, and (
maximum anticipated pressure surge forces were considered in developing l design loads for the RCS makeup system. The staff considers these to be acceptable load considerations for this c,1 plication. The staf f has not
! reautred the 9:5 pressure control and makeup system be evaluated for seismic load capability. .
- 3) Other Considerations j
Welded construction will be used wherever possible to minimize
! the potential for system leakage. Component s will be fabricated from stainless steel or carbon steel clad with stainless steel.
, 4) Evaluation Conclusion ) We conclude that the licensee has specified components that will
- be designed and fabricated in accordance with acceptable Industry codes I
or standards and will taken into account the loads associated with i i
! startup, testing, and e=pected system operation.
l 1 The use of components that are in conformance with these clrteria l provides adequate assurance that structural Integrity of the Reactor Coolant Pressure Control System will be maintained. a i i { I
C. Structural Deslon The system includes the follmaing major equipment: a) Three water tanks, 900 gal, capacity. Each weigh 20 kips empty and 27.5 klps when full of water. Each tank will be supported on four 12 In. by 12 In. plates. 5 Borated water tank weighinc 63 kips supported 71 four 6 In. plates. c) Two - 100 HP pumps weighing 5.4 klos each. The enclosed Figure 2 (2 sheets) show the conditions of the original structure. The area is located between columns AP and AT in south-north direction and colu,ns A65 and A67 in east-west direction. The slab Is three feet thick and the reinforcing steel 15 #1 at 9 In. top 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 yleid stress of the reinforcing steel is 60 ksi. The licensee has analyzed the slab for the additional loads resulting from the new equipment and concluded that the stresses will be within the allowables using generally accepted codes. The analysis was performed for the static conditions only. We conclude that the licensee has performed his analysis in accordance with the methods and procedures which are soecified by the aonropriate codes and standards. The use o' these methods provide reasonable assurance that the structures aected by this moji'Ication will continue to Der #Sem their Intended
'urctlan.
Y----
l p ,g ( 5 y j) 9,. l~ M V EL 313.f'j WW ' . . .
' o = l- R '
W. O . 2 5 I 1' RCS PQstsoRE. courQoc WS 4 g
? YWM No.*diso
.. .F r _h;:. 4 :s* h _.. wo- h '
w-r.__ .h. A ,I i -
, e.' 1s' _ .. h *.'
e au , TJ
- e, m , . u . . . . mt .,, T.
.... . y
\. .
8 . *
.* \ .
. .,f- - ; ;
.. - *.. 'e .
I ' _m,., - i
..,. tV l -
'?
i_.. . v N yj f *4 I i c
,._.w..
N~ . 4 . u iQ i
/,w co Joe " t' h,. .
. . . , ;' . .. . c/ cs comwwi
\ w. yalm_ .+ .
pg . it~ &
- F.
v g ~() ;
'--~.-1 iii' j; ggm.,
. u . . ..m $, .
- i n
4 .
\ ' ' ; ~
g v }; v . .. :. m--- --
- .~ T --
... / '"*'"'.;; "
;i i
- p.1,
.!n,,._ g...
g 192,w - . -t.. 4 \ (T - TV l is-L o
~
. . . . . . . . . . .i . . \ .,,, , , ,
.. ... r .t.
La. . i. ,q .. . ,, .e_r , m .
./c .sy..
., ,;), aa
.. 3 ;;. . ...
I u. g ._ _. _.M _ i. , yl .s.t f -35t L. ,F - JA.. _ _ _ _._ _ . , .s .o a W/P ,g ; ,, ' b:l , J_ -- f,,o.,, _f_t'1.% l w,((/y,bl, t ret ' '._.' 11 'if l ai J. - 3p, _p_ '" ~
- p. J * .
~g--- .-- -
,.,.....s. /y
. . _ * , - - .1 ,s
,,j , ,
,,. i g : ,h m,' '"
?
.-..u...
.__n_r_. .,, l r u. ...
. l ,, . . . . . .. _n.s .- ~
.t_]
, n au 1-,j
%, .p,. r_&_ j, r'r. jy-gp2-r --l-
- v. ..-..i_ l.,. , c. _______ __ d~'<.ti
*r'- ,' - - - s Jr- - - - -
r.
~
y___. -l-+ hy 1 3 a
=___F--- - -- ..
I e
.w. .i. w.,.a-
/ \7 i \'nM.,l! -
i l M.m
,,-.m
'l
. J, s..<.. ..
', ,,, j . 5 .*
l .i t
. si y
~.
- n. ;.\ '!) \ t ..
, . . ,i _ svat ; . . . .:e ,
[~' .. 1. mo, . i
' 4
.'L fL-__;. o 1 . _ _ .. , _ _ . ,
. ._ . . l . . .Q
_. i
,* I j i i v
,,.a, . e . . ..
;e
;i r .....- i e
'._e..
,,,, g l - - . t
. .r i y, ..,2_. . ........ . . err
,j ' '
y :'q,:, ......... ...... _.1. ._
/) .
,,..as. . .... . . _ _ _
[ .. Q v ...
..y,
], ,l,l l
i l .... I,, ',
,4, ..
. . _ - ..y ..
l ,
....__.- .. .. ..: q
_ . . . . .. . j . .$q,u. .__ _'_'....,l hII "* . i, l lll l ,, a..,, __ ___g_ _ _ _ a
'a . . . .-
l, 4 l tn g
>444
..r., j l 3 i
.; I .
j.l i j ) L_,..
, n4 'I l 4 '
s gp.
...s.
.. g .-
'., l.4. .. ,.. . . . 4,spa__.
s ,,.. ... . .. I
.... L. . . .
_a.......__. ___.m__
. . . .. _ }. . i. Q-ti , gi : .! rr i 4
g : > h15 p ,3. A (skaf'd ' i . 8 m ib i y $
.e ;
Il y *gt,'i g 8 W i en -e ;
,3 t a les a33 w -
D i. i
, . s. .
- i . s. 8 1
? v s./ :-
,1 1,f, k-2
$s$ j > '/
n- . su
- 5 t Y:- ',g;iir:s. =
3._'i
#1~.__ .
. i a3 x g
.. ,. , .- ., co 1 ., .
3 f ~el j
} @
=
i, .3 n . - en s d5 :
.d
. . t . .
. :t
,:r- t -
u.r$ .
- 9, .._.. .,s .
] ..
. I'.. f
- w a "-
..J.s
.... 3g .1-i emmaa l e,!!. 9. 3 A, e. e-i 6
gg - 3 /; ' :) _1
-2
. d .; , 1.
e (.. ) 3--- ti=
- i
; g.;g.
. s t.r(
u .-. [r g g gg :- E.s. 2. __, T. N-
- g -
- 11, 1 y, ht
,i. F, ;
i
.k .
(3)- hg > s .
' l s
J - . i
.u @ .; '
u
;: c , .- .{o m s-y,
,_ 2 5 1
a,I :: = s
- 3 c g1 3'r . .. e s
% - i - --- . j
; . ,. . . . , .r. e ,
l. i- . t o Lc< . t . . e :- 1 II :- i -
.. :1
.r.
s
,i .
lrtg5 i W.. h p 1 ",.. ' e q ; 7=y-{.
.t
$ fp._3 7,a--
9
.\ l !.4 ,u d.r - - ,L_,..r.._
, . . .- de:
a > i(f
} ,.
I,5 ) If
\
> E3 ti '
[63
, Lm. - - i
- ::: e -
fi ist
. l I ,
i 1 j
- 12 D. In-trumentation and Control i
The following Instrumentation is to be provided for the Reactor i Coolant Pressure Control System: i l 1. System Vater Pressure
- 2. Vater Level In each tank 3 Nitrogen Pressure I 4 Make-up Flow I
S. Make-up Pump Discharge Pressure
- 6. Borated Vater Storage Tank Level
- 7. Sarated Vater supply Temperature i Redundancy is provided for: Water level (In each tank) and system water ;
4 Flow. Nitrogen pressure is sensed at the cylinders and just downstream I i of the N2 pressure regulators. , i j The design of the system is not predicated on the availability of . the pressurizer instrumentation and controls. The ranges of f l Instrumentation have not been provided. 1 The control aspects of this system are as follows: The first make-up i pu o will se cut in by a " Law" level signal frem the nitrogen pressurized 1 !, surge tank; the second pump wIll be cut i n b y a " Low L ow" level signall j and both oumos will be cut out by a "High" level signal. The controls l
)
1 for the beaters have not been described. l
?
l I i . 1
.m. y%.- ..----.-,----,-e --~.-----s-- .--,-w.w- .--- . e- - -_.3 , . . . . ,,-----,y---.--.,.,-----,,-c--,. .m, - ,,m--, -,w-
I i We 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 1 i the motor operated Isolation valve initiated by low level in the tank i nearest the Reactor Coolant System. The Isolation valve will be signaled i to close when water in the tank is deoleted to 300 gallons. Tank pressure
. taps will be utilized for level Indication. (One pressure tap in the I tank and the other in the 6. Inch piping upstream of the tank). Because ,
i of the location of pressure taps and uncertainty of level control to actuate the isolation valve under transient conditions we require functional testing 1 of this control system that would require tank blowdown prior to Installa-I' tion of the pressure control and makeup system. If the test results turn i out to be unsatisfactory other means of preventing N2 Insurge into the I RCS such as automatic isolation and venting of N2 supply will have to i be considered. We have further required that an alarm be provided to the operator when the di'ferential pressure of the Reactor Coolant 9 J ! System and the pressure control system exceeds a set value. The licensee has complied and the alarm will be annunciated whenever the differential ) I' pressure is greater than 50 psi. We find the instrumentation and control I . aspects of this system as described above to be acceptable, i e 4 i l I-t l'
- _ _ _ _ _ _ _ - _ _ -_ =- _ . _ - _ , _ _ - _ _ - ___ - _ _ .. - - -
4 l 1 i 34 .
- b. Decav Heat Removal Svstem (Existino and Skid-Mnunted)
)
Oceay 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 alditional i skid mounted train of decay heat removal equipr1ent, and 3) considerati9n 1 -
of a dedicated long term decay heat removal and primary coolant cleanup system in a permanent structure. I A. Unoradino E IstIno Decay Heat Remnval System Provisions will be made for conducting a preoperational test of each loop of the existing OHR system. Locations of system leakage will
; be identified using television cameras Installed at key locations. Once 4
leakage paths are identified, they will be corrected if possible, thereby providing as leak tight i system as is practical. Leakage collection cacability will also be added to the system where feasible (i.e., collection of leakage around valves). Instrumentation to detect OHR purp vibration will also be installed. The availability of the existing decay heat removal system will depend on the final leak tightness of the System considering the contamination levels present in the primary coolant. 1 i i i i 1 i
!s -
S. Dasion of New Skid Mounted Oecay Heat Removal Svetem A third train for decay heat removal will be provided. This will 1 Involve a tie into the existing decay heat removal system drop line downstream of valve DH-V-3 located in the fuel handling building and ties 19to the two return lines to the cold len- 'also located in the fuel
~
nandling building (See Ficare 1). New lines will be run through the l nenetration room to an opening cut in the fuel Fandling building wall 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
'ull capacity punns. The discharge line for the heat exchancer will return through the openinq in the fuel handling Sulldinq to the return l
l line tie-Ins. The tie into the decay heat removal drop line will be made by welding an 1 Inch weldalet to the pipe with a full penetrati on I i weld, dye penetration testInq the weld, then cutting the hole In the olpe using a clasma are cutting process to minimite debris and finally 4 we'dino the rew pipe to tFe weldolet. A similar procedure would be used i for the tie-Ins to the two return lines. This procedure should minimlre the time that the decay heat removal system will be out of operation. All valves will be electric motor operated and will have two seals with 1
. p r o i l s l ona. for collecting leakage and directing It to the existing rad-waste system. Additional concettlons will be proviled In the new piping outside the fuel handling b'illding 'or 'uture u e in the Installation o' a deditated, long li'e. ba rdamed f t ruc tare whl e will contai n bea t I
i l I L 1 e i l !
' 1 d exchangers, pur'ps, deminerali. i ers and filters for long tern decay hea t removal and cleanup of pr* nary coolant water. The secondary side of the new decav heat renovat heat exchanger will be cooled by a new separate decay heat closed cooling water systen with This syste, i.a tarn its wn cumo, oiping and valves. (See Ficure 2).
,i l l be e, ole: Lv water -o, the nuclear services rive- water s'* tem.
i Nes conaectinns will be made to this systen, i The design of this new decay heat removal systen will be conoatible I with the current primary systen decay heat levels and operating parareters, it sill require that the pressurizer level be maintained half full at all times to insure adequate NPSH to the OH4 pumps. Afternativelv, the backun makeuo and pressure control system could s s rve to satisfy this requirement. The new skid mounted DHR syste ,and closed : nling syste is cocol?telv S parate and Indenent from the e.<:sti v M :ut is not fesigned again t single active 'altare. Hwever, ! t*e existinq 3HR systen would be available in this event. i i j ' The olan is to have the skid availaole but not make the connection to l the lines oenetrating the fuel handling building wall at this time. i In the event this sytem would be re 4uired for immediate service, the 4Wid ani clolog would be ehleided to provide all reas7nabIV ac"Iovable f reJiation oeitect:$n. t
--r --_wr-, - - - , .- . * , -,,,-,-- - - - , - , + =- ..-_1N~ -,.o -7 - - - . - - - . - - - _ wy.,,.y-_ _. --6.-w. ----wmi ~_. ,-_m - i...-g___-..-, y -- _,, .. , -- .i -..e
@s.
{! ti e s " o. __. ,?_ . . . _. h,AI.._,_d_i._
> y > u .. . . . -.
3g
- a
[)_ , [h_ $.k.d- __ __. . M S I
.3g. ..
\ ed.J_._ r,V_x,..Lm we
. ..$_ b.
.... W .
. w-i _, y - y
.a ke
' ~ '
f O 9 E = a b
~
% , -- G - - __ Q
-. -..._.. -- g . _ -
-g -
Q _
._f3
. . . . . .M . _ ._ ____. _ . _ .. _
k._ _ _ . _ . _.. ._. w 2 __.____ y 7 q .. e>. . _ _ _ _ _. . _ _ . _ _ .
~ ._ . Q __.
. _.___ 4 _ _.._. _. $ -._ _ . - ._ . . . . . .
% a
_. _ _ . .__ _ m . _ _ _ _.
/ T ,
.,.3 ___ .~ . _. . . . . _
K _j - _ w. A4_ _ _ . _ ._. _- _.* _ .3 5 - - 2 a . _. W ._ _. M A A L _ .. . . _ - _ -. . . . _ . _ . _ . . . .
# .- .t
.y
. A Q *-
__...._- _ i _ __ __ _ *, ._ _ t . .a . _ - . _ . . 1 Q . I .._.
- a 1 .
.y . . _ _
$Nb ,.. ' i . .. .
....---..._.c 4 ._... _ . . - . . . .
T . -_4 g.4 : - . . . - - _ _ . . - h _. .. nB . .; /i _ g -. - _ . . . . . - _ . - . 4 w g 7 k7 eg: b _ _. dgg,. t y >H Ul - . _ . . . _ . c
.w . _ . . _ . _ . . . _ . _ . _ . . . _ . . . .
a .
., g N ,' -
-> 42 .< s
; .._ _ f (A c,y o p. Y * 'NCL 3 i! e 2,
o otoe. 9 60G. g g e . o 4. o A ~ . -
- .{ e 3*c
$Eu p. . . . rd.4 . _ _ _
u .v
;'- q esc3a
-(,(-0# V44 u c a3 y. . . - m - - --
.x n m
A3 _. s . _ g _
. ~
, ; .r . i r '
i ldCvd trkiTitJGIloEs !cde/l'/ tir/LY : gg; oceny;; c u>st )'
- i. ;
- -- i D . Coo ttidG I Wr17C# C '/ .< TCM i ,
I i j : i
}{ CAT! CootER l ,
i i 3 ! i
! I r- ~ ~ !-~ ' i I l
PCFCR ,T0 fFIG. l l, BCCAf f 4 ' - _L _L,,% 71 l h} % 4_ - j ! I l l , l' W l llCAT ' [ ~. i l j r - 3 60th'/illrvr ctosto
~
j : ; i: t ' ,,
, l jq- n!grWA.'"'f iSCIIt I ' '
i - - - i j t ! I I coatiWG - l varu supac j u_ 1_, @eyv7!coottri i g- ; k
- --i i is .
l l ,
- e l qorv, ! :
3; W ' raim !
, - i . . >
' j' l l. 1 i
- : ,a ! rr in.. i j j ! i bbP i j
j l
.t" l i j 'i ;
i ;E I I y '- ' l i ed_v; i I l i , O' j ';i i ,'l' ! [SHALbmu. i i j l i
'}
i i l i i l Ro ,
- b o
=
l f=' ors'u$$' ; l Edt>Cch',HN)YCLoshp 'N' ; i ; , i ! ; ; ; "C'WMCR.Pvty I i i
. l t I I 4 l ' ! ; i j i ; j i i ! : i .
r I. i lt i
. i i I !
!! ! I fuTtsBC occM l!
- i I .
l t
' 3
!! !.'s i l
i ' i l l i i i
, I i
l HEM w&Tce2C(O sys7(nSCO
. ; Co l
l I li I l 1 ; ! SlitCLb ,$tL; ; >
,, revg ;
I T'f, fi- T -
- l. l f 3 I
l lI - %l\ filt:Ib WAIL' ,, p i - , I gg.g H gg j i
,, Ji i.,,- t w _w. ;
i 17I - t , f
. . i , :
i l ' 7 7 l i o ; m f
+ + NEE bCC'l/ , ilEAT !
j iCAVicE .ccctE R ' V d ' y ' fctrarsu6teir)/ j g i ntd i c'est o d'suik l l : uu sas .- i, ! , i ! i !. . . l t. ! ! i i i . , i ., Wart e ( m..avy g t
; ---+ - -
gyw p i, .evisrinG uscitAal ; l g; ,,,57,,,a ,,,,e t e,,f $79 4,e,t _g} . ,
- - -> y-S s'cRJ ! cts 'StVt/ l . 1 s ty ; ; s
peqca s,p,sgp --wp, ._ A .3
$-. .- ' sys. I i
! l tystit R. ,
ii; j l , . i l ; billol line s'- ! l' l FIGLtIQE !2i ! r el'er eutt i:sg(tles& olFj.) or erl s'i h 3 i i ' O
The staf f has reviewed the expected system performance of this system and concludes that its heat removal capacity is su'ficient for decay heat removal from the TMl-2 reactor. Mechanical Deslon Considerations The skid mounted decay heat removal system consists of components such as Feat exchangers, purps, tanks, valves and piping. The applicable design codes and standards used far the design o' these components are provided in the table below. , Table i APPLICABLE DESIGN CODES OR STANDARDS Pumo and Heat Evehancer ASME Section 111 Cl.2 Valves ASME Section lit Cl.1 Picino Mixture of Type 304 and 316 10" Sch, 40 Stainless Steel sections. ASTN Material Certification All welds fully radiographed except for weld-o-let connections to existing DHR piping. I 8 1
, i .i
. -- - - . - . . - - - - - - - -- ._ .- ~ -.. - .. - . - .-- - _ -
1
. f l
I Deslon Celteria
! Loads Considered - Normal operating conditions - to include operating i j t j pressure, dead w-Ight, pump vibration, thermal e 3 t
{' expansion, and maximum anticipated pressure surges. [ i l' j Normal stress limits will be met for all piping and . components including l I i loads from OBE. l l Design Stress Limits Used: f s Pumps, Valves, Heat Exchangers - as specified in ASME, Section lli for I applicable Code Class.
; i i Piping - Stress limits per ANSI 8-31.7. ; .i i Desion Information? Sneelfic for Svstem Tle-In to Existina DHR Piping, , ! Weld-o-let Connectinn '
j Reinforcement area of fitting provides a 240 percent margin over the area of the existing DHR piping it replaces. 1 i , i j Pipe supports will be arranged so maximum stress levels at the weld- : 1, . j o-let to OHR pipe interaction will be held to about one third of the I 1 8-31.7 allowable stress limit for normal loads, i j . I i i j Weld-o-let to OHR existing pipe walds will be made using a qualified ; I ! procedure and by welders qualified on weld-o-let to pipe connection } f mockuos, t t j !
=
d i i i j - l _ . _ . _ . _ _ . . _ . _ _ _ - , . . -., . .,_ -_ .- ,__ _ __ _._u;,._.__-._-._..-...._ - _ __,- - ,_ _ ,_ . _
. .- - - . _ ~ - . _= . . - . . - - -- . -_ .. .. . - - - .- - -
j . l Because of time constraints, the weld-o-let to pipe welds will not l be radiographed; however, the root of the weld will be ground and dye j ] penetrate inspected and the final surface will be dye penetrant Inspected. I
- Additionally, the design Is being quallfled by hydrostatic pressure tests
! and bending moment tests which apply loads until the simulated existing r
OHR oine exceeds its vield strength. t -
! All we' ids and the cut into the existing DHR pipe will be performed i
4 using the clasma arc method. The plasma arc was chosen for the com-bination of small heat-af fected zone and mininum resulting slag which 1 can be cleaned up with relative ease. 1 i 1 ) Miscellaneous Valves - Line valves and relief valves will have leakage or discharge
} fluid piped to a draln tank In the auxiliary building.
i i 3*cav Heat Closed CoolIno tlater Sv< tem Components ASME Section 111 CL.3 For all components I i Materlats: C.S. piping l S.S. Pump, Valves, Heat Exchanger i
- All connections welded except for piping to cmponent Interfaces which 1
j will be flanged.
! i f
1 i
. , _ . - . - . _ . . - . _ . . - _ . - s - . . - . - _ , _ . , . . . . _ - _ , . . . . _ . _ , . . . . _ _ _ _ - , _ - - . . - - , _ _ _ .
m___.___-. __.m.,,,,__._,_
.i
- LO -
d-Evaluation
Conclusion:
We have concluded that the Licenece has specified components designed i and fabricated in accordance with acceptable lodustry codes or standards and will take into account the loads associated, with startup testing, and planned system operation, h The use of components that are in conformance with these criteria provides adequate assurance that structural integrity of the Decay , h Heat Removal System will be maintained.
- L i
). Structural Fuel Handlino Buildino Vall Penetration (Aux. Bldo) I In order to provide the skid mounted deda,y,' heat removal system a 5 I' penetration would have to be made through/'he t west wall of the Fuel Handling Bldg., between column lines AC and AF and across column line i: A69, at elevation 297'-0" (See B&R Drawing 2075). This is approximately seven feet below grade. Excavation outside of the structure will be done by pick and shovel to minimize a possibility of danacing any L piping or electrical conduit. The outside wall at that location is reinforced concrete, 5'0" thick. The compressive strength of concrete In this wall is 5000 psig. e
- l ii.
a di %
41 - The opening is to be made 21" wide 5'1" high on the outside tapering to 4'-1" at the inside face. One reinforcing bar, #18 is to be cut at each face. Figure 3 shows the front view and the cross-section at the opening, Procedure At the time o' writing of this report various methods of penetration are : being studied. One possibility is using the oxygen lance process. A trial penetration is currentiv being performed by this method. Drilling through the well i s also considered. The final method of the wall penetration will be described upon pending results of the test and con-sultation with the other experts in the field of construction materials, a Instrumentation for New Decav Heat Remnval 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. I The following instrumentation will be provided: J I 1. DHR pump suction pressure
~ '
- 2. DHR pump suction temperature (DHR cooler inlet temperature)
- 3. DHR Cooler inlet pressure f 4 DHR Cooler discharge pressure J
i 5. DHR Cooler discharge temperature
- 6. DHR flowrate a
i i . { ! l 4 mg *y tw rm- 9,- ----a- +simmu-er w -p ----- e tw >= -s m e w-, e --p y- ~e+g4 -P - w-W-sms-a w+rp*---r-e- - - - -p
L - 7 gS7;Yl w 4 ' AI / , l '- 9 " J u : ; ,f /'l' o , il , g,g,;c,q I I insid 'I' .';'.
,. .,. o' ,
I l,, ., 7
]l l
l Exisi #l2 reinf
+o be cui Each -
.fa c e . .
t' -
, _i 3 %
i i i 1 E/. 2 9 5 L 5'V] V' T I ils ;, s- ,, Ai s , , 3 , s ,f , , s i s 4 s, -
, ,, l ' s
( 5 L o" m i-Fronf Vie vi Seciton A-A f~/GURE 3 Skc ch Show/co
~Penefm?!on H, rough i Auxilia-.ru 3!c/c> ,
!A!.n/l.
9/upn.
- _ _ , _ _ - . - ...-,y,..- . . - . , - _
Redundant measurements and display are provided for each parameter and separated to be consistent with the A or B pump train. Independent power supplies are provided for each train of instrumentation. No automatic control or interlock features are provided. To the extent oractical, the nuclear inst rumentat ion which is sa fety
. related has been purchased to Class IE requirements. The instrumentation selected was based on, to the extent available, the same manufacturer, model, and principle as the instruments used in the existing DHR systen.
The staff concludes that the instrumentatinn and controls to be orovided for the skid mounted decay heat system are appropriate for their intended function. O k
. . _ _ _ . _ _ _ _ m . - _ _ _ _ _ =_. . . . _ _ _ . _ . . _ _ _ _.._. _ __.
=
1 1
. h3 _ -
i
- 5. Electrical Power System Modifications A. General Description i
The licensee has identified a number of modifications to the Unit 2 power system in order to acconmodate a loss of offsite power. Backup power to the two existing safety related , load groups (designated Red 1 and Green) is supplied by the two existing Class IE diesel generators. No loading changes have been proposed for these busses and the staff position is that all new loads be powered from other busses leaving the j original Class IE system Intact. The staff has. further stipulated that i the loading of these Class IE busses be on a " manual only" basis due to i the time available for operator action following any perterbation and the complex nature of trying to coordinate which loads are needed and which would be detrimental if actuated for each potential plant mode l of cooling. Approved written emergency procedures are required to i cover the above contingencies. 'Je find the above described modifications ] to the existing Class IE system to be acceptable. For all electrical loads that previously did not require loss-of-oFfsite power back-up protection and all newly added loads which ! now require this protection, the licensee is installing two new j, diesel generators. These diesel generators will be assigned to busses 2-3 (breaker position 3-3) and 2-4 (breaker position 4-13). ' i l
These load grnups are now designated " Gray" and "ilhite" respectively. There diesels are self-contained skid-nounted outdoor units rated nominally at 2500 kw each. The convention 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 modifications. Ne have required that the new load
- grouns be mair.tainec incecendent and that no interlocks shal' be incorpo-atej :nto the p2ver system design that couples the lead greaps in any .Ta nn e r .
Each n?9 diesel will have i t s own switchgear, control battery, and fuel tank installed adjacent to it. The switchgear will be connected (vie underground cabling) to basses 2-3 and 2-4 (gray and white reseectively) through the existing circuit breakers (previously ysed for condensate booster numpe Co-p-2A and Co-p-2B) with nodified relaying to acconnodate the oever sources. Loss of of# site power will be detected by new undervoltage schemes which will automatically start the diesels. The breakers at busses 2-? and 2-4 will be normally closed and the new diesel generator circuit breakers will be aut>natically closed when the units - have attained rated soced and voltage. Motor loads are automatically tripoed upon loss of power as part of the existing design. The staff requires that these busses be manually loaded by anproved written procedure. Thes2 actions can be cerformed f rom the control roon. The basis 'or this reovire ent is the time available for operator action and the di'Ferent loads requi ed 'or tre var.ious potential plant conditions. The licensee's design is in con #er ance with this requirement and is acceotable.
. hg . .
Busses 2-3 and 2-4 have existing bus transfer schemes that allow con-tinuity of power supply, should one o' the two offsite power circuits through the auxiliary transformers be lont, by fast-transfer to the other transformer. This scheme is to be left intact and the new undervoltage
. detection schemes will have a 10 second delay to accommodate the fast transfer if one circuit remains available. Ve find the above - moi;#ications to be acceotable.
We will require that test ing requirement of tFe Gray and White diesel generators and their associated 0.C. control power supplies (batteries) be comparable to those that now apply to the Class IE diesel generators and batteries. l S. Stean Generator A and 3 Short Term Solld Vater Operation i The 900 horsepower condensate pumps are the single largest loads that will receive diesel generator back-up power source protection. The circulating water pumos rated at 2250 horsepower each have placed an additional restraint on the power system. This size load is too large for any of the four diesel generators now on site (red, green, gray and white). 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 pumo while supplying power to the first. Cooling system requirements are 'ulfilled by one pump operation.
4
\
i < 4 h6 - , i 4 . i , I The new 13.2 kv offsite transmission line will originate at the Middletown Junction Substation. This substation has a 230 kv and a
.l tl i' 115 kv bus tied together by four auto-transformers. This modification t utilizes auto-transformer no. 2 and connects to the 13.2 kv tertiary t
winding rated at 25 MvA. The group operated air break switches on the '1 j l. 230 kv side will be coened to isolate the trans'er from a direct I connection to the 230 kv* system. There is a spare 2]O kv line that j terminates on the tower at the entrance to the MI'ddletown switchyard , on one end and terminates on the tower at the entrance to the Three Mlle I t island Switchyard on the other end. This line shares a double-circuit l; I ' t line of transnission towers with an existing 230 kv line to the plant. ) At the Middletown Junction Substation, a short line section from the : i existing capacitor breaker on the 13.2 kv tertiary of the no. 2 transformer back to the spare 230 kv conductors must be constructed. The capacitor I i
- bank will be disconnected. The necessary additional relaying will :
be provided for this caoacitor breaker to protect this new feed. j At the Three Mile Island Substation, a 13.2 kv underground cable h -
' supply from the spare 230 kv conductors will be run around the I southern side of the natural draft cooling tower. This underground
. I portion of the cable run protects the line from any of the other Once around the cooling tower, the line i: ' incoming lines falling on it.
kv trans- l
- goes overhead for one span and terminates on a 13 MVA 13.2/4.16 I t
former. A spare 10 MVA transformer will also be in place with manual switching l f i, +r t I i . I
capability for energiratien should the first one fall. These trans-
'ormers are located adjacent to the circulating water pump house and the l
! electrical switchgear bay. The 4 kv portion of the line ties into breaker cubicle 5-4 on bus 2-5 located within the switchgear bay. This breaker 1 was originally used to suoply power to circulating water pump CW-P-lE. . . This breaker will have modified relaving to acc7mmodate the new cower i source. We have requireo this alternate power source scheme to be
" manual-only" by approved wri tten procedure. The bus must be cleared of all connections following loss of offsite power prior to reenergization and subsequent loading of the circulating water pump. There is a normally 4
open bus tie between busses 2-5 and 2-6. Closing this bus tie gives access to five of the original six pumps. 1
- We will recuire that the 13.2 kv line be tested weekly by energizing i busses 2-5 and 2-7 for a short time interval to assure continued
} l functional capability. There are no phase angle dif ferences between I the existing syste,and this new line so that the connection may be i made on a live bus followed by tripping circuit breaker 29-52 to prevent tieing the 115 kv system to the 230 kv system through the plant distribution system. We further require that circuit breakers T-56-2 and I - T-71-2 be verified open or a daily basis. l J t
. , . - - , , , , ,- . , . . , . ~ . _ - , , , - - , , , - - - - . . _ . , - - - ,,- . . , . - - - , . . - - - - - - - , - - - ,
,_ . .. - . ~ _ _ _ _ -- . - _ - - . . . . . - - - . --.
l
. 41 . ,
- 1) the aoplication is limited to the circulating water cumps.
- 2) these pumps are only required for the time window necessary to complete the modifications to steam generators A and 8 'or 'ong l -
term core cooling.
- 3) the time required to have the 13.2 kv line operational is consistent with the requirement to orovide back-up power as soon as practicable.
- 4) alternatite propo als such as the use o' diesel cenerators were act considered feasible due to the si.:e requi rements for starting 2250 horsepower loads and indoor service would require a new building that could not be built in time based upon existing schedules. '
i t-
- 5) the new line hae acen isolated to the extent practicable from the l 230 kv system and therefore may survive a local 230 kv system disturbance.
- 6) given a total grid blackout, there are six combustion turbines I in the cloce proximity of Three Mlle Island. These uni ts are rated 23 MVA each and have black-start remote supervisory I control. The system dispatcher in York has supervis3ry control 4
of this system with the one exception that the system dispatcher in Lebanon would have to be consulted as to the oosition of
~
certain intervening circuit breakers.
- 7) transmission line tower failure could not redner the 13.2 kv and the 230 kv systems inoperable.
t
-. - . =_. .- _ _ _ __
. 49 -
We find the addition cf the 13.2 kv transmission line as a backup power source to be acceptable given the operational needs and time restraints present at Three Mile Island Unit 2. C. Steam Generator A Modifications These modifications provide a new 700 horsepower high pressare pump for circulating water through the secondary side of the A stean gererator. This new loop is in cooled 5v the Nuclear Services River Vater Systen and the Nuclear Services River Vater Purges. The new high pressure pump will be powered from bus 2-3 (Gray), the Nuclear Services River Vater Pumps are power:d from the existing Class IE power system. Ve i find the power source allocations to be consistent with the separation of i the A and B steam generator nodifications, to capable of supplying the practical requirements of the system and to therefore be acceptable. l
- 0. Stean Generator B Hodifications These modifications provide a new 700 horsepower high oressure pump for circulating coo 1Ing water through the secondary side of the B steam generator. This new loop is in turn cooled by the Secondary Services Closed Cooling Water Systems which utilizes the Secondary Closed Cooling Water Pumps. The final couting loop uses the Nuclear Services River Water Pumps for circulation. The new high pressure I
pump will be powered f rom bus 2-4 (white), the se-ondary service closed cooling water pumps !C-P-IS and SC-P-IC are powered f rom 490 motor j control center (MCC) 2-418 which in turn connects to 430 voit bus 2-4I i
.50 .
and through a 4160/480 volt to bus 2-4 This arrangement allows a back-up pump (one purnp operation meets cooling system requirements) and is part of the white power system. The Nuclear 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-4 ments of keeping the B steam generator cooling svstem on the even numbered separate busses, to be capable of supplying the functional requirements i l cf the system and to therefore be acceptable. E. Skid-Mounted Decay Heat Removal Svstem This new system will have 480 vo't motor operated valves .i arranged in such a manner that there will be two sets of isolation valves o, 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 capability 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 400 horsepower pumos. These valves and the associated pumps will be powered f rom separate busses (Gray and White) to assure system function given a single power source failure.
~
Two new 480 volt motor control centers will also be provided. All electric oowered equipment (valves, pumps, motor control centers and cabling) vill be Class IE system quality. 3ecause of the low decay i heat levels, suffi: lent time is available 'or aanual operator action assuming loss of o>ter or equipment.
The secondary cooling loop for the DHR heat exchanger includes an additional 250 horsepower pumo connected to 410 volt bus 2-44 All associated motor coerated valves will receive normal power from the white power system. Should the white power system fall, backup power can be orovided ' rom the gray system by closing the normally open bus tie benaker between buses 2-44 and 2-3h. We require this tie l 1 "r'aker be racked-out at all II9es ard only c?csad upon the ' allure of the white power systen by acproved written procedure. Tnis system will not be used concurrently with the steam generator cooling modes describes above and there# ore does not affect diesel generator capacity. We find the elect rical power aspects of this design as descriced above to be acceptable. F. Reactor Coolant Pressure Control System All electrical equipnent and instrumentation required to j operate the system are powered from the gray and white 90wer systems. The charging pumps (A and B) are rated 100 horseoower and are povered
. from 420 volt notor control centers 2-32A (gray) and 2-42A (white) respectively. The charging water storage tank heater is rated at 100 kw . and will be powered f rom Sas 2-45 There are a number of snall loads associated with this system that have not been assigned power sources.
I
l
. 62 . .
The system will be aut. mated as soon as pnssibic but this nav not occur before initial operation. A motor operated isolation valve will be provided to automatically close when the water level T. the tank nearest the reactor coolant systen is approximately one third full. This is to preclude the introduction of nitrogen into the reactor r .
;v i ::ola-t 5yste,.
t i ll o c
J _ 53
- 6. Quality Assurance Program vor TMI 2 System Modifications The NRC Technical Review Group has reviewed and evaluated the Quality Assurance (QA) Program of GPU/ Meted and of their major subcontractors for the TMl-2 system modifications. These QA Programs recognize the uniqueness of the THI-2 plant condition and balance the schedule urgency for c0maleti3n o# svstem Todifications agaInst the extent of i
a:of i:ation o' traditional (A crocran p actices consistent with ( maintaining assurance that soecified system requirement- are met. 1 The GPU/ Meted QA Program has been specifically tailored for the TMl-2 system modi fications. The Program will apply QA criteria of 10 CFR Aependix B commensurate with the specified system recuirements and will be comoatible with the Meted Operations QA Program previously i accepted by NRC. GPU/ Meted has established a QA organization at the TMI-2 site specifically responsible for the system modification (A activities. This staf f is excerlenced in all QA disciplines and associated t:chnical fields, including mechanical, electrical and civil engineering as w 11 as welding and non-desctructive examination. The GPU/ Meted QA Manager and QA engineers were brought in from the Forked River facility. GPU/ Meted is the lead responsible QA organization for the TMl-2 system modification program. Their QA Program will provide surveillance over the activities of their subcontractors, including Vestinghouse and Burns & Roe. Jestinghouse has established a QA Progra, to control CA activities associated with design, pro-r l
~
{
- 3h -
curement and vendor component fabrication related to the skid mounted Backuo DHR system modification. Vestinghouse has organized an experienced QA staf f specifically responsible for this task. Vestinghou=e will also provide system installation and pre-operational testing procedures and on site technical supervision which will be conducted subject t, the 3PU/ Meted 1A Program. Burns & Roe is responsiole 'or establishinq designs '7r the other TMl-2 system modiri:ati9ns, including the designs for the Reactor Coolant P essure Control System, the A and B Steam Generator CooIIng Svsten. Burns and Roe is implementing design control RA practices which assure that appropriate quality standards are specified and included in design documents that provide for verifying or checking the adequacy of design and that will control design changes. The GPU/ Meted QA Program will orovide 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 staf f experienced in mechanical, electrical and civil engineering disciplines and welding and non-destructive examination to provide surveillance of system modification activities at TMI 2 site and at equipment vendor facilities as necessary. 1
?
3e . Based on cur review and evaluation of the CA nractices, :ontrols, and organi::ation of GPU/ Meted and their major subcontractors, we conclude that these QA Programs will assure meeting the criteria of 10 CFR 50 Appendix B commensurate with the TMI-2 system modification requirements and are acceptable. 9 9 L i}}