ML19220B150
| ML19220B150 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/15/1977 |
| From: | Herbein J Metropolitan Edison Co |
| To: | Silver H Office of Nuclear Reactor Regulation |
| References | |
| GQL-1129, NUDOCS 7904250481 | |
| Download: ML19220B150 (21) | |
Text
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METROPOLI FAN EDISON COMPANY TELEPHCNE 215 - 329 2001 PCST OFFICE BOX 542 RE AclNG. PENNSYLV ANI A 19603
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Oirector of :inc. ear Reactor Regdatien a y ^*.,.,
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Attn :
Mr. Harley Silver Civisica of Reactor licensing
$*g Qj, Q in2 db>{,7-Q/
U. S. :Iuclear Pegulatory Connissien y
y' Washington, D. C.
20535 i
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Cear Sir:
Ihree '." ' a 7sland ;iuclear Stttien Unit 2 ('"'C-2 )
license !!c. CPFldo rocket :so. 50 ;23 FSAE Amendment ??c. S Infomation to expedite your review, enclosed are advanced ccpies of pages which N
Aa-
- !o. 58 to the we anticipete vill be included with submission of Amendment T:E-2 FSA3. 2ese pages address the Stsff's concerns on pre-cperaticnal re-spense time testing and docu=ent cur response to these concerns.
Shculd you have any questiens, please centac:
re.
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,r truly yours,
/
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/7. G. Herbein Vice - President
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s v.a Inclosures: Fares 1kA-53 Supplement 3 bl.23/22.29 7.0,15.0 (;uestien and Eespcuse, 5 pa es) s 3.10-a
.1 m.,. c,.
D2 7904250 181.
225 *ec.;2_,
.1 33-22-29f 33-22-29s 33-22-29g(i) 33-22-29h S3-22-291 Table 22.29-6 (Continued) S3-22-2?j Figure 15.1.1-9
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.w Table 153 k 15.1-21 (d
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TF 330/h 1.
Purecse 1.1 To verify preper operatien of circuits and ccupenents which receive input signals free the Red Drive Centrol Panel.
1.2 To verify proper dash;ct Operatica by de=enstrating red drop time censistency.
2.
F~erecuisites 2.1 Reacter fuel not leaded.
2.2 Each CRDM has been filled and vented.
2.3 TP 330/3 Centrol Red Drive Functienal Test.
2.h Inter ediate eccling must be available te all CRDM's.
3.
Test "ethed 3.1 Apply initid pcver to the CFIM's frc= the centrol system.
3.2 Cperate CRD syctes frc= the Red Drive Centrol Panel.
h.
Data Recuired h.1 Red positions to check for correct syste Icgie operation.
h.2 Chart receriing of red positien versus tine.
5 Accentance Criteria 5.1 CRD systen cperates as specified in ranufacturer's instructica manual.
5.2 Insertica time frem trip initiation to full insertien should be 2.+ 0.2 seconds.
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Supple =ent 3 41.23/22.29, 7.0, 15.0
- 1. There are several secticas of your FSAR that appear to have incon-sistencies and/or errors relating to respcase tires for resistance temperature detectors (RID's) used in the reactor protection Specifically, Table 7.2-2 lists RID response times as systen.
3.40 seccads; Table 3.10-2 lists a required RID respcnse ti e of 5.0 seconds and a typical tested value of 5.25 seconds; Table 22.29-5 lists an PTD component specification of 5.0 seconds ; and Table 22.29-1 lists RTD sensor response tire as 8.0 seconds.
In additica, Revisien 3 of BAW-10003A, which is referenced by ;mur application, states that each production RID is tested to ensure it has a time constant of 5 seconds or less and that this ti e constant is deemed acceptable frc= a safety analysis standpoint.
Table 22.29-5 states that no values were assured in the safety analyscs,since no transients are ter=innted by this trip paraceter.
However, Table 15.1-1 states that: (1) the che ical and voluce control systet talfunction would be ter=inated by a reactor trip on high ecolant temperature or pressure and (2) loss of nor:al feedwater would result in a reactor trip on high reactor coolant temperature or pressure. Also, Table 7.2-1 states that the =rxi-rate of change (Accident) for the reactor coolant outlet te -
mu:
perature detectors 14 the Single Red Group Sta-tup Accident.
Resolve the apparent errors identified above or alternatively, provide explanations for the differences, icur response should also include a clear descriptien of the design basis for the RTD response tones; define the testing methods to be e=plcyed in ed deter =ining response times and explain the correlations of as response times under test conditions with required response t under actual service conditices.
- 2. Table 22.29-5 lists assu ptiens used in the safety analyses for individual channels of the reactor protection systen that appear inconsist ent and non-conservative relative to Figure 15.1.1-9 in your FSAR. The error appears to relate to assu=ptiens used for breaker opening and rod release time. Modify either Table 22.29-5 and/or Figure 15.1.1-9 to show that safety analyses assumptions provided in Table 22.29-5 are equal to or conservative with re-spect to information provided in Section 15.
If Figure 15.1.1-9 is =cdified, state how control rod scra: tire require =ents will be affected.
Also, the safety analyses assu=ption of 650/1400 :s for reactor coolant flou listed in Table 22.29-5 is non-conservative relative to a stated response tira of 650 s provided in Table 15.1.5-2.
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Supplement 3 Modify year FSAR or provide a technical argn=ent to resclve this apparent error.
In your response you should clearly state if additional delay rire due to snubbers was u ad in your safety analyses and provide the basis for the 1400 ts listed in Table 22.29-5.
- 3. The required Ingineered Safety Features Response times and their bases provided in Table 22.29-3, 22.29-4, and 22.29-6 appear incocplete, inconsistent, and/or non-conservative with respect to each other and/or other infor=ation provided in your FSAR.
Specifically, the following areas need correction, clarification, or justification:
a.
Response ti=es and bases are not specified for the auxiliary feedwater syst2=.
Provide required re-sponse ti=es and their bases.
b.
Ite= 2.c (2) in Table 22.29-3 and Ite=s 2.c. (2) (a) and (b) in Table 22.29-4 and the associated bases provided in Table 22.29-6 are not totally consistent with values provided in Section 6 or your TSAR (Table 6.2-15).
Clarify the response tices and bases in the above tables to establish that response times of individual contain=ent isolation valves will be consistent with those stated in Table 6.2-15. In addition, valve No. MU-V25 (listed in Tabics 22.29-3 and 22.29-4) is not listed in Table o.2-15.
State the function of this valve and resolve or explain the incensistency between tables, c.
Ite= 2.e in Tables 22.29-3 and 22.29-4 lists response ti=e for contain=ent cocling as 6 125 seconds and
- 114 seconds respectively. The stated basis for this respense time is the fan cooler delay assuced in the stea= line break analysis provided in Table 153-2.
It shneld be noted that your design basis LOCA analy-sis provided in Sectica 6 of your FSAR assuces a core li=iting value of 35 seconds for f an cooler operations.
This sa:e analysis assu=es that reactor building spray vill be operating within 95 seconds folicwing the LOCA.
Resolve or explain these inconsistencies, d.
Ites 2.f (2) in Table 22.29-3 lists a response ti=e of 95 seconds for the Nuclear Services Closed Ccoling System.
The basis for this limit thac is provided in Table 22.29 states, in part, that the Reactor Srilding E=ergency Cooling Booster pumps start 100 1.
F
=
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Supple ent 3 seconds after the accident; therefore, there is a 5-second =argin batween the establish ent of Nucient Services Closed Ccoling water for the booster purps and motors and the tire the booster pu ps start.
Also, Ites 2.g in Table 22.29-3 lists a response tine for the Nuclear Services River Water Syste of 95 seconds. In view of the LOCA analyses provided in Sectien 6 of your FSAR vhich assure the availability and operation of the contain=ent fan coil units within 35 seconds and that these fan coil units are supplied cooling water from the Nuclear Services River Water Systen via th7 i eactor Building Energency Cooling Booster pu=ps, the proposed response times of 6 95 seconds for Items 2.f(2) (Nuclear Services Closed Cooling) and 2.g (Nuclear Services River Water) appear non-conservative relative to assumptions used in the LCCA analyses. Also, the startup time for the Reactor Building Energency Cooling 3 coster pumps of 100 seconds does not appear consistent with LOCA accident assu=p-ions. Resolve or explain these ac mrent
.inconsistencie,
e.
Items 5.b (1), (2), (3) and (4) in Table 22.19-3 list response tices for Feedvater Isolation that are non-conservative with respect to the assumptions used in analyning the stean line break accident provided in Table 153-4.
Resolve or explain this apparent incensistency.
Response
- 1. In order to clarify the inconsistencies noted above, the following table is presentcd listing the old (pre-ISAR) and nev (present Safety Analysis) temperature response tire specifications:
RTD Plus Bridge RTD Only RTD Plus Bridge Plus RPS Delay Old Specification (177GY) 3.00sec 3.25sec
- 3. f.O s e c New Specification (177'dW) 5.00sec 5.25sec 5.40sec
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Supple:ent 3 Tables 7.2-2, 3.10-2 end 22.29-1 have been revised to eli=inate incensistencies. Table 22.29-5 and 3AU-10CO3, Rev 3 are correct as written.
- The high temperature trip has not been used in any of the Cupter 15 accident analysis. The high temperature trip has been.
luded in the RPS as a prudent precaution to ter=inate any transient which causes the hot leg temperature to continually increase. As a backup trip,it is not needed nor taken credit for in the analysis,
rad Table 15.1-1 had previously been revised to delete reference to the high temperature trip.
Table 7.2-1, which lists maximus rates of change of various plant paraceters, does not have any direct analytical connection to the design basis for the RTD.
Since the high tenperature trip is not used for pri=ary protection of transients, it is not required to track the temperature rise reported in Table 7.2-1 and this table has been = edified to reflect this fact. It should be e=phasized that the R D is designed to caxinua accuracy for steady state protection and not for card =us speed of response.
The control
- 2. Figure 15.1.1-9 is in error and has been revised.
rod scra= time require =ents are based on the analysis and correctly reported in Table 22.29-5.
The value given in Figure 15.1.1-9 is in error and was not used in the safety analysis, therefore, no change is necessary to the control rod scran ti=e requiretant.
Table 22.29-5 has been revised (see footnote 3) to explain the apparent discrepancy in the use of the snubbered vs the un-snubbered instrument response tice.
- 3. a.
Tables 22.29-3, 22.29-4, and 22.29-6 have been revised to provide the response time and bases for the
- crgency feed-water system.
b.
Tables 22.29-3 and 22.29-4 establish the required response ti=es for contain=ent isolation valves. As such Table 6.2-15 vas revised in acendment 57 by deleting the ti=es listed on that table. Valve HU-V25 is listed on Table 6.2-15 at pene-tration K-545D; it is the reactor coolant pucp seal return line.
Supple =ent 3 Tables 22.29-3 an'd 22.29-4 have been revised to show the design c.
basis LOCA as the basis for centaintent cooling, d.
Refer to ites c. abcve. Also note that the 5-second =argin between the establishment of NSCCW for the booster pump toters and the starting of these cotors ceets both the design re-quire ents for the cotor and the required accident response tite.
Table 22.T_9-3 lists a respense ti=e for the feed regulating e.
valve of 9.2 seconds, which includes a 6.6 second stroke ti=c for the valve and a 2.6 second feedvater latching delay. The accident analysis of Appendix 153 assucas a valve stroke ti=e of 4.4 sec. and a feedvater latching delay of 2.6 sec. The 4.4 sec. tine is tha. valve closure time frca the position re-quired to pass 100% feedvater flov (70% open). The valve closure tine is tested frea full open; hevever, the closure ti=e from the 70% open pcsition is verified to be less than 4.4 sec.
Sectic: 153.3.3.4-d has been clarified to crplain the diff erence in respense tites.
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i T_ABLE 3.10-2 SfENDOR TIME CONSTANTS _
Typiertl Requir ed Plant Varinble Tested Value Value Simulntion Sennor 25 ms 220 en Decrenne Preneure Presnure (MotoroIn)
Increnne Temp?rnture 11 3 25 nec.
3' Tempernture
- 3 0 nee.
l o.
5 25 cec. ****
83 ITT GY 5 0 nec.
177 IN 110 mn 7
250 mn N
Decrenne Flow 210 r.c h Reactor Coolant Flov 2h0 ms Lons of Pumpo N Renetor Coolant Pump Monitor **
Less than 100 ma ##
u 250 ms Increase Preneure C Renctor Pullding Pressure""
(Barton) flow raten of 50 to 80 feet per d
- R*ID'n are required to have a responne time of less than 5.0 secon a atAll RTD's with matched b ve 600 F.
The acceptance teotn are conducted
,necond at operating temperaturen a ofactory in laboratory testing.
An annlytical extrapointion of for acceptable renponse time at the f 3 feet per necond.
a unit tempernture of 120 F with n fluid flow rnte o for the Rosemount 177 UW RTD and thermowell 0
the RTD laboratory time conotant resultn to operating plant conditionsto achieve n tim conota l t operation conditions, at Higher opernting temperaturen b
t y tenting.
m ot achieve a time conntant of 6.8 secondn or louer in in ora ortime constant under in-ne indient.cn that h
or flow rnten would renuit in a further reduction in t e
- Provides n digital (on-off) sigunl.
'"" Manufacturers information.
h
- Includen 0.25 nec. time conotant for bridge.
L Table 7.2-1 I::KP: ATIO:: COI:CEPl!I;;G GEilEPATII:G ETATIO : VARIABLES TO BE FONITORED TO PPOVIDE PROTFCTIVE ACTIO::
- 'arrir Eet=ce: ?ruie -.
m Paximun Rate Prudent Gpersticr.al Li.i*.s a i o
{
Measurement of Ch inge Purnps Operational C r.s e t cf it.:sfe
--id le Frrors JAccident)
Operating Limit Cc.rii' i..:
f, s
A.ga.
T.:
Out-cf-Core 6.5%
h805/sec Four 0 to 1005 1 125 ir-power lecel m
- .. u.~.n ?l ux (0.65%/Ak/k
(?twer Level)
Rod Ejection)
Three O to 755 121.65 in pcwer le' -
h Two (One Pump 0 to h9%
1 26.2I ir. pcv;r
'e...
Each Loop) 100%
1 21 i n flov
'c. PC Flow Technical 25%/see Four Specification (Locked Rotor "inimun Values or Loss of Threc 7h.7%
> 27. 'i i r, fi c.-
Accumed Coolant Flow) h%
> h.1E in fic "
Two (One Pump 9
o Each Loop)
J t-c.
?? O-ict Ter.pelature IF No t. used for Four 532 to 60h F 1 11.5 transient analysis Three 532 to 60h F
> 20.8 Two (One Pump 532 to 60h 7
> 33 5 Each Loop) c.
.;., Prencure 30 pai 100 psi /nec Four 2200 psig
> P0t prir icv.
(Sin 6 e Rod
_ 550.0 pair
.s'r 1
Group Stirtup Three 2200 psig
> 320 e ir 1cw, 7;,
Accident) 550.C sic.ich Tao (One Purip 2200 psig
_ 500 pair 1r,
Each Loop) 55L.C psi; i.ig'r N(g "i'-3 Ccrre latien limit, not DNB limit H
y 4
. Se.
.: '. '.2 Ji r...[:..
.".3
.......t.
.. 'N......
.. ~
i i......
.f 4
Table 7.2-2 DF.ETOR PmTrECTIO:t fiYSTFN Titlp GErrI!C I.lMITG One Denetor Coolant Piunp Four fienetor Coolant Pumps Three penetor Coolant Pumps Operating in Each Inop Operating (Nominni Operating (Nominni (Nominni Operating Snutdown Desponse operating Power - 100%)
Operating Power - 75%)
Power L91) lvpnsn Tiren. see g
g 1.
Noelenr power Max. %
105.5 105 5 105.5 5.0(3) 0.16 e
or rnted power ti ne
?.
Nuclei power te ed on 1.07 timen flow m' nun 1.0T t imes riov minus 1.07 times rio minun pypassed 0.L9 N
'loWI' nnd (mlinInnce, reduct{on due t0 reduction due tO reduction due to mar, or rated power imbninnee(s) imbalance (s) imbninne e( n )
1.
Nuclenr power ed nn NA (Gee note 5)
NA (Gee Note 5) 77% S)
Dypanned 0 39 I
pump moni tor n,
mnx. $
of rated power la.
liigh revlor coolant 2355 2355 2355 1020I4) 0.LO atyr. tem prenpure, pnig, mnx.
r1 5.
Inw ren: tor coolnnt 1900 1900 1900 Dypanned 0.bo f,
nyritem presnure, psig, min.
b 6.
Varinble low reactor (16.25 Tout-703's )III (16.25 Tout-78 % )(1 )
(16.25 Tout-700)
D Pa" Sed 0 k0 7
coolant rystem pressure psig, min.
7 Benetor cooinnt temp.
619 619 619 619 5.400) l r.. un..
8.
High Renetor Nullding to le le L
0.LO prennure, pntg. max.
(1) T In in der.rceo Fnhrenheit (F) o u t.
(?) Renct or cooinnt nyntem riow. 5
( 3) Mministrntively controlled reduction net only during renetor shutdown g ',
( 8: ) hier.atica.lly r.-t vhan et her acaaw nts or the 3011 (nn noccl rled) are byp.w.ed el (5) 1%e pump monitorn nino produce n trlP on (n) loss or two re actor coolnnt pumpn in one s eni L..r coolnnt loop, and (b) loan or one or two reactor coolnnt pumps during two-pump operati m.
(6) Pump monitorn itallente the loss or a renetor coolant pump when the mennured lever to the pump is equal to or less than b
PM or t he runninc 3.ower.
3 70 U) Response tien includes sensor (5.00 sec.), bridge (0.25 see), and RPS (0.15 sec.) response times.
I N
a
i TAllIE 22.29-1 SAFETY HELATED SET 1GORS MAX MFH SETIGOR set! GOR MODEL TYPE GYSTEf4 TIME (SEC.)
SPEC.(GEC.)
ITEM PARAfETER _
1.
Ileaci,or Coolant 11 nil?y HY D/P RPG 0.25/1.0 (5)
Transmitter Flow 2.
Henetor Coolant seatine, house Prenn ure RPS 0.220 0.10 n
Veritrak 59Pil Trancmitter a
Prennure 0.10 g@,
3 Itenetor Building Darton Pren o ure HPG 0.250 288 Gwitch Prennure g-1 Is.
itcactor Coolant fuchenter Wtts HPG 0.2 40 m
Tranntlucer/
Pump Power Innt.
g 111 n te,le m
7 5
Reactor Coolant Foxboro Prennure GFAG (1)
N Prencure Ell-Gli Transmitter d
6.
Henetor Du11 ding Barton Prennure GFAG (2) 0.10 Prennure la pal 288 Gwitch 7
Henetor Building United Elec.
Prennure Reactor 1.0 Prennure-30 pni J 30;-156 Switch Building Spray Static "0" Ring Prennure Feedwnter
( 3) 8.
Main Steam Prennure 9fi-AAl5-CSGX Gwitch Intch 9
Henc tor Coolant Ronemount RTD HPG 5 0 (i) l i
[
l Temperature 17TIIW (Table 22.29-2, Item 3) nuat be f,2.0 sec.
(1) Sum of nennor, llistnble ntring (Table 22.29-2, Item.:) nnd Logic Relnyn
/
(2) Gum o r nennor and Logic Relnyn (Tisble 22.29-2, Item 3) nunt be 61.0 sec.
( 3) Sum of nennor and Lor,1c Relayo (Tnble 22.29-2 Item 14) nunt be 6 2.6 sec.
W I
(la) Tented orroite by RTD vendor (also see Tablen 3.10-2 and 7.2-2).for sensor with snubb-. Gennor and unnbber will b (5) Time npec. In 0.25 sec. for senoor alone,1.0 sec.
together i f applienble.
n
O Supple =ent 3 Initiating Sirnal and Function Restense Tire in Secends h.
Feneter crelant Pressure - Icv a.
High Pressure Injecticn 5.25*/25**
1.
Lev Pressure Injectica 1.25*/25**
c.
Cc=penent Ccoling Water System (1) Decry Heat Closed Cooling
- f. 3c0 * /300 * *
(2) Nuclear Services Clesed Cooling f,95*/N.A.**
d.
Service Water Syste=
(Nuclear Services River Water) 1 95*/95**
5 Steam Gene-ster Pressure - Lev a.
Main Stea= Isclation 1 N.A.*/12k.6**
b.
Feet ater Isolation (1) FV-V30A/3 i N.A.*/9 2**
(2) FW-V17A/3 1 N.A.*/32.6**
(3) FW-725A/3 1 N.A.*/lk.6**
lhI FW-Vl9A/3 1 N.A.*/32.6**
i 6.
E=ergency Feedwater Pu=ps a.
Motor driven 29 */12 **
b.
Stea= driven 29/NA**
The ti=es listed above are the allowable ti=e intervals fro = when the cenitored paraneter exceeds its ESF actuation setpoint at the channel senser until the ESF equipcent is perfor=ing its safety functior Diesel generator starting and sequence leading delays included.
- Diesel generator starting and sequence leading delays not included.
Offsite pcVer available.
S3-22-29f k'
- =
supplenent 3
".'A31E 22. 29 h T'4'G".Fr..rs" c A "...".v W*
".'.m..W A
I.'CIVILUAL SI2CCS Functien Resnens e "'ine in Seconds 1.
- a. Diesel generator start and connect to bus (ES signal) _ill
- b. Diesel generator start ced cennect to bus (LCOP signal)il7 2.
- a. Eigh Pressure Injectica 1 lh
- b. Lov Pressure Injecticn I lh
- c. (1) Centai--a + " urge Isclatica 1 3*
(2) (a) Phase I Centa#--- + 'sclatien valves 1 59*
(b) Phase II Centainment Isclatica valves Ih9
( 3).YJ-725 1 6h
- d. (1) Centrol Rec: Isolatica 1 5*
(2) Centrcl Building EVAC 1 869
- e. Cente.inment Cooling I llh
- f. (1) Decay Heat C1csed Ccoling i 289 (2) ?'uclear Services Clcsed Ccoling 1 8h
- g. Nuclear Services River '4 ate-
< 8k
- h. (1) ES-V1A/S 1 12 (2) DH-76A/3 1 25 3
Centainment Spray Pu=ps 1 20 k.
- a. Main Stes: Iselation Valves 1 120*
- b. (1) 74-V30A/S 1 6.6*
(2) Pa*-717A/B 1 30*
( 3) Fa'-725A/3 1 12*
(h) Fa'-Vl9A/3 1 30
- S3-22-29g i3-295
Supplement 3 5.
- a. McCor driven c: ergency Feedwater purp fl2
- b. Steam driven e=ergency Feedwater purp 129 a
- Diesel generator start time does not apply to these subsyste=.s.
S 3-22-29 g (i) 73
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0 ee l l dt n N W r
0 mm a
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o 0
0 0
0
)
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s 1
2 4
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m P
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TABLE 22.29-6 BASES FOR ESF RESpCNSE TDES LISTED IN TA3LE 22.29-3 2.a.
FSAR Section 6. 3. 3.13 2.b.
FSAR Section 6. 3. 3.13 2.c.
(1) NRC Standard Reviev ?lan 6.2.4, page 6.2.4-11 (2) NRC Standard Review Plan 6.2.4, page 6.2.4-7 (3) This single isolation valve has a closing time in excess of one minute when the diesel generator start time is included. The valve is in series with a second isolation valve in the same line with a closing time of less than one minute. Since the line is in a closed syste=, built to ASME Code Section III standards and seis=ically supported, the effects of this slow valve closure are considered to have an insignificant effect on the overall leakage fro the contain=ent following a LOCA.
2.d.
(1) Same isolation criteria as used for chlorine. See FSAR Supplement 3, page S3-310-3.
(2) Mini =u: time to reach limiting equiptent temperature.
See FSAR Table 7.3-3.
2.e.
Question 3.4, Supple =ent 2 represents the design basis for the i
ccatainment cooling response time.
The analysis assu=es that contain:2nt cooling is available in 125 sec.
2.f.
(1) Decay Heat Closed Cooling Syste= is initially required to provide cooling water to the Decay Heat Removal (LPI) pump and cotor coolers. The pumps and motors supplied are designed to run at runout flow for five minutes with co cooling water supplied fro the Decay Heat Closed Coolfng System.
(2) The Nuclear Services Closed Cooling Syste is initially required to provide cooling water to the Building Spray pump and motor coolers, the Makeup pu=p and motor coolers and the Reactor Building Emergency Cooling Booster pump and totor coolers. The Building Spray and Makeup pumps and motors are designed to run for five minutes with no cooling water flow. The Reactor Building Estr"3ency Cooling Booster pu=ps start 100 seconds after Ehe accident. This provides a 5 sacond cargin after cooling flow is established to the punps and motors, assuming loss of power has occurred.
For the case with no loss of power. the Nuclear Services Closed Cooling pumps continue to run uninterrupted.
Therefore, a response time requirement is not applicable for this case.
2.g.
The Nuclear Services River L'ater Syste initially provides water to the suction of the Reactor Building Emergency
(
Cooling Booster pumps. The 5 second cargin before the booster pumps start assures adequate suction pressure for these pucps prior to starting.
S3-22-291 N ~ D$
TA3LE 22.29-6 (CONTINUED) 3ASE FOR ESF RESPCSSE TDES LISTED IN TA3LE 22.29-3 2.h.
(1) This time is based on expected opening time of the valves, and is conservative with respect to the required ti=e of 31 seconds, when the systen is to be at full flow.
(2) This time is based on expected opening time of the valves. The valves begin opening prior to Building Spray pump start, and at 31 seconds (full spray flow) are 40% open. This is satisfactory for spray system per formance.
3.a.
This fa the required time for full pu=p flow, and is the basir used for determining the 95 second response ti e (FSA'i Section 6.2.1.1.5) for full spray flow frc= the spray no::.les.
4.a.
Sar a as 2.a.
4.b.
Ss:e as 2.b.
4.c.
S a:e as 2. f.
4.d.
Sa:e as 2.g.
- 5.a.
FS AR Appendix 153, Table 153-4, Low stea=line pressure feedwater latching setpoint reached (0.4 seconds) and Main steam isolation valves close (125 seconds).
- 5.b.
FSAR Appendix 153 Table 153-3, and Section 153. 3. 3. 4. d),
Assumed feedwater latching delay time of 2.6 seconds.
6.a.
Supplcment 2, Q21.43 (conservatively assuming offsite power is lost at the time of reactor trip with energency feedvater initiated by a loss of all RCF's ) and Section 15.1.8.
6.b.
Section 15.1.8.
- No response time is given for loss of offsite power case.
See FSAR Section 153.2.4 S3-22-29j 73 299
(
iuu f
/
/
/
90
/
/
/
E 80 C
S5 70 c
a 60 E
a:
50 40 E
O=
30
(
-ce 1
2 20
/
10 0
c 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0
- 2. 2
- 2. 4 Time after CROM release,* seconds
'I hk, du ROD WORTH INSERTION VERSUS TIME THREE 1.!!LE ISLAND NUCLEAR STATION UNIT 2 AD7CS/
p; @ FIGURE 15.1.1-9
<s dh
e TA3LE 153-3
(
VALVE ACTUATCR DATA CLOSURE TDE CONTROL FROM 100". OPEN VALVE SIGNAL (SEC.)
W-V30A/B - Main f eed regulating 1,3
- 6. 6
- W-V14A/3 - Main feed upstream block 4
90 W-V17A/3 - Main feed downstreas block 3
30 W-V25A/3 - Startup 1,3 12 W-V26A/3 - Startup upstrea= block 5
90 W-Vl9A/3 - Startup downstress block 3
30 Turbine S*.op Valves 2
0.5 MSlV-7A/S, 4A/3 3
116 1.
ICS Signal 2.
Turbine Trip 3.
Feedvater Latch Signal 4.
50': Closure Position of W-V25 5.
Re=ote Manual Only Although cicsure time of the =ain feed regulating valve is 6.6 seconds. a value of 4.4 seconds is used in the ar.alysis. This is the closure ti=e of the valve from its positien corresponding to 100%
feedwater flow.
c<,d
7 (,1 153-29
IA3LE IfB-4 CERCNCLCGICAL SECUENCE OF EVE'iTS FOR STEAM l':2 ERLU" *a*IT'4 TUR3I:2 STOP '7AL'l: FAILURE - CCNTAI'i:'ENT/:!ODERA~I
- OVERCCOLING CASE Time Af ter Event Ruoture, Sec.
Double-ended rupture of sain steam line 0.0 Lew stea=line pressure (600 psia) feedvater latching setpoint reached 0.4 High building pressure (4 psig) ESFAS setpoint reached 1.4 Emergency feedwater injection begins 2.0 Reactor trips en variable low pressure 3.0 Main feedwater control valves begin to close 3.0 Centrol rods begin to drop; turbine trips but valve fails to close 3.5 Peak fission pcuer occars 3.6 Startup control valve and feedwater block valves begin to close 5.0 Main feedwater control valves closed 7.8 *
{
l Main feedwater startup control valves closed 11.0 Low RC coolant pressure (1600 paia) ESFAS se tpoint reached 13.0 High pressure inj ection begins 24.5 High-high building pressure (30 psig) ESFAS setpoint reached 19.8 Main f eedvater block valves closed 35.0 Core flooding begins 41.5 Mini:eu suberitical =argin occurs 43.0 Building spray begins 46.0 Peak containment pressure occurs 83.8 Reactor Building spray renches full capacity (3000 gpc) 91.0 Main stea= isolation valves close 125 Building f an coolers tura on 126.4
- Assuning closure of the valve fron position corresponding to 100% Feedwat2r Flew.
O t.
f L' equipment are assumed in the analysis if they produce core serious con-sequences) er cperator action. Operater setica for maintaining het shutdevn ecnditicas or ecoldern to ec'd shutdevn conditions is assumed indicaticas l to be accceplished only where adequate tire and instrument tre available to the operater.
i Figure 15.1.1-9 shows the curve of narcaliced rod worth versus time l
i that was utilized in the analysis of each accident which results in reactor trip. Transient analysis only takes credit for the rod worth g
necessary to overecte the te perature effects fro: hot full power to 9-hot zero power and to provide a 1% ak/k shutdown =argin, including the highest worth rod stuck out of the core. In no case is the rod worth required greater than the worth shown on Figure 15.1.1-9.
Table 4.3-10 shows that the =ini=u: net rod worth ave.ilt.ble for the case (end of equilibriu= cycle) is 1.4% ak/k assu=ing a stuck worst red worth of 2.2" ak/k. It can be seen frc: Figute 15.1.1-10 that g
af ter only 75% travel of the control rods, 80% ef the available rod werth will have been inserted. Even if all control rods were assumed to n
stop at their 75; insertion point, 80" of the available red worth (verst case) would still allow a shatdown tarzin of abcut 0.5% in this unlikely set of circu stances. The basis for nelecticn of the trip reactivity insertion curve is described in 3AW-1461 " Reactivity insertien Assunptiens Used For Safety Analysis Calculations."
Heutren pcver is used in this chapter to describe the total er.ergy re-lease frc: fission. Therral pcVer refers to the total core heat trans-fer rate frc the fuel to the ecolant. Reactor ecolant syste: pressure is used syncnyrcusly with the core outlet pressure. A trip setpcint designates a cere ccnditien inclusive of =aximu: calibratica and instru-centation errers which cust be exceeded bercre reacter trip is assured to occur.
The safety evaluatica criterien which is adepted in these accident anal-yses to insure that the reactor coolant systen boundary integrity is
=aintained is that the syste: pressure shall re=ain belev cede pressure licits. The AS"E Ccde Sectica III pressure licit is 1105 of the reactor coolant systen design pressure, 25C0 psig (see Chapter 5). The safety li=it thus established is 2750 psig.
The safety evaluatien criterien which is adopted in these accident anal-yses to insure that no fuel damage cccurs is that a ON3R greater than 1.3, as predicted by the W-3 correlatien, =ust be =aintained throughcut the transient. As denenstrated in Chapter L, a ONER of greater than 1.3 exists for the 1125 design overpever ccnditien. Thus, if the core ther-
=al power during
-"a
- -a sient dces not exceed 1125, there vill be no fuel damage, except for the locked roter analysis in 15.1.5 If the DNER gces telev 1.3 during a transient, the gap activity for all of the fuel rods with a ONER cf less than 1.3 is assured to be released.
If 9
15.1-2 mto ely
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a- = 'a. - 1 3
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- e -
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- 16. An add'*4-..=.'
'"4 e i da==3 a..'.a...'w..
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- -d ria
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7"e.' a.....'.j*.' a ' *. >. *e-="se e-'
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.as*..a e." ene.gy a--.4..'....
^ r '..". e - ^ 4 a i a. *. *..i..7 a c.ide.*., *'e fuel integrity is :aintained if the peak enthalpy of the hottest rod is i
, less than 210 cal /g=, the threshcld energy for the :irecnia -vater re-action. Above 210 cal /g= the next thresnold is 260 cal /s=, above which the fuel rod vill probably n:- be intact. The " safety =argin" for any transient is the difference be:veen the peak talue of the centrolling para eter and the 112% ther=al pcver,1.3 ::i3R, 210 cal /g= or 2750 psig i'.. safety li=it.
--- - ~ - - - - - ---
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15.1-2a b* {} /~
9 r '- r ?
I LJ