Text

Proposed Tech Specs Reflecting Rev to Plant MSSV Lift Setting Tolerances
	ML20083B635
Person / Time
Site:	Prairie Island
Issue date:	05/04/1995
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20083B627	List: ML20083B625; ML20083B635;
References
	NUDOCS 9505120145
	Download: ML20083B635 (49)
	v • d • e

i l

- 1 LICENSE AMENDMENT REQUEST DATED May 4, 1995 Pressurizer Safety Valves and Main Steam Safety Valves Lift l Setting Tolerance Chanze and Safety Limit Curves Channes !

EXHIBIT B Appendix A, Technical Specification Pages Marked Up Pages ;

l TS-1 TS-viii l TS-x TS-xiii TS.2.1-1 TS.2.2-1 TS.2.3-2 TS.2.3-3 Figure TS.2.1-1 (old)

Figure TS.2.1-1 (new)

TS.3.4-1 Table TS.4.1 2A (Page 1 of 2) .

Table TS.4.1-2A (Page 2 of 2) (new) l TS.6.4-1 B.2.1-1 B.2.1-2 B.2.1-3 (new) i B.2.1-4 (new)

B.2.1-5 (new) ,

Figure B.2.1-1 (new)

B.2.2-1 (new)

B.3.1-2 B.3.1-3 B.3.4-1 B.3.4-2 (new) j I

l 9505120145 950504 PDR ADOL',K 05000282 P PDR l

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

TS-1

??? 91 10/27/99 TECHNICAL SPECIFICATIONS l l

IABLE OF CONTENTS

, TS SECTION TITLE PAGE ,

1.0 DEFINITIONS TS.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING TS.2.1-1 ,

2.1 Safety Limity, "rreter C::: TS.2.1-1 2.2 Safety Limit 91'61 tim m ?, Pereter Cecirrt TS.2.1-1 "y: t er " r r r d$ ' ' ~~

2.3 Limiting Safety Systeu Settings, Protective Instrumentation TS.2.3-1 A. Protective Instrumentation Settings for Reactor Trip TS.2.3-1 .

B. Protective Instrumentation Settings for Reactor Trip Interlocks TS.2.3-4 ,

C. Control Rod Withdrawal Stops TS.2.3-4 ,

i 4

1 d

- , _ . , ., m _ . _ .

TS-viii REV 95 5/9/91 TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1 6.1 Organization TS.6.1-1 6.2 Review and Audit TS.6.2-1 A. Safety Audit Committee (SAC) TS.6.2-1

1. Membership TS.6.2-1

2. Qualifications TS.6.2-1

3. Meeting Frequency TS.6.2-2

4. Quorum TS.6.2-2

5. Responsibilities TS.6.2-2

6. Audit TS.6.2-3

7. Authority TS.6.2-4

8. Records TS.6.2-4

9. Procedures TS.6.2-4 B. Operations Committee (OC) TS.6.2-5

1. Membership TS.6.2-5

2. Meeting Frequency TS.6.2-5

3. Quorum TS.6.2-5

4. Responsibilities TS.6.2-5

5. Authority TS.6.2-6

6. Records TS.6.2-6

7. Procedures TS.6.2-6 C. Maintenance Procedures TS.6.2-7 6.3 Special Inspections and Audits TS.6.3-1 6.4 DnisEed S fety Limit VI:12 tier TS . E . ' 1 6.5 ElahE"dperating Procedures TS.6.5-1 A. Plant Operations TS.6.5-1 B. Radiological TS.6.5-1 C. Maintenance and Test TS.6.5-3 D. Process Control Program (PCP) TS.6.5-3 E. Offsite Dose Calculation Manual (ODCM) TS.6.5-4 F. Securtly TS.6.5-4 G. Temporary Changes to Procedures TS.6.5-4 6.6 Plant Operating Records TS.6.6-1 A. Records Retained for Five Years TS.6.6-1 B. Records Retained for the Life of the Plant TS.6.6-1 1

l l

I o l 1

1 TS-x REV 9^ 3/20/91 TABLE OF CONTENTS (continued)

TS BASES SECTION TITLE EAG,5 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits. Ecceter Core B.2.1-1 DIREin~df6FC6fsTSsfitfitisiEEffMP"N""ESTERI

((gjAh[Rjjyj$Cd611h(S,yajjp[gyyyfe[S{f[5$$$[MQ2Bj$Mj 2.2 Safety Limit yiplatipns, Re cter Ceelant B.2.2-1 Sycter Pressure ,

2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B,3.4 1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 B. Gaseous Effluents B.3.9-2 C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 E. & F. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Control Rod Operability Limitations B.3.10-9 H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 ,

3.12 Snubbers B.3.12-1 3.13 control Room Air Treatment System B.3.13-1 3.14 Fire Detection and Protection Systems B.3.14-1 3.15 Event Monitoring Instrumentation B.3.15-1

I

~

TS-xiii

?_"? 1^* 9/3/93 t

APPENDIX A TECHNICAL SPECIFICATIONS ;

LIST OF FIGURES

.TS FIGURE TITLE l

2.1-1 S:fety L! nite , Reactor Core M], 50- ' "ydr rli l 5: L::; ^;:retier 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit j Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 1 3.8-1 Spent Fuel Pool Unrestricted Region Minimus Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid-Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Caseous Effluents 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 shield Building Design In-Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Region Minimum Burnup Requirements 572?I:1

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TS.2.3-2 REV 91 10/27/99 4 2.3.A.2.d Cont, and f (AI) is a function of the indicated difference between top l and bottom detectors of the power-range nuclear ion chamber, with gains to be selected based on measured instrument response during '
plant startup tests, such that wgere qyt and qb are the percent power in the top and bottom halves of the core, respectively, and qs + q3 is total core power in per- cent of rated power: ;
i
~~1. for qe - q3 within'-12% and g-94, f (AI) - 0, and~~
~~2. for each percent that the magnitude of qs - q3 exceeds i~~
~~+94 the AT trip set point shall be automatically reduced by an equivalent of 2.5 percent of RATED THERMAL POWER.~~
~~3. for each percent that the magnitude of qt - qu exceeds~~
~~-124, the T trip set point-shall be automatically reduced by an equivalent of 1.5 percent of RATED THERMAL POWER.~~
~~e. Overpower A T !~~
K 33 t sT .
ATp s AT, tr4 - - K6 (T-T') - t (am l
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T - Average temperature, 'F T' -
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K4 s 1.10 Ks
- 0.0275 for increasing T; O for decreasing T K, - 0.002 for T > T', O for T < T' ta - 10 see f(AI) - as defined in d. above .
~~f. Low reactor coolant flow per loop - 290% of normal .~~
indicated loop flow as measured at loop elbow tap. !
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~~1. Reactor coolant pump bus undervoltage -~~
~~E75% of normal voltage.~~
~~2. Reactor coolant pump bus underfrequency -~~
~~258.2 Hz~~
~~h. Power range neutron flux rate.~~
~~1. Positive rate - s15% of RATED THERMAL POWER with a time constant 22 seconds~~
~~2. Negative rate - 57% of RATED THERMAL POWER with a time constant 22 seconds~~
~~3. Other reactor trips~~
~~a. High pressurizer water level - 590% of narrow range instrument span.~~
~~b. Low-low steam generator water level - 25% of narrow range instrument span,~~
~~c. Turbine Generator trip~~
~~1. Turbine stop valve indicators - closed~~
~~2. Low auto stop oil pressure - 245 psig~~
~~d. Safety injection - See Specification 3.5~~
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- " - - ..:... - " -. -... . . . ". .- . ". ." _' . . . _ . ."l...._ _.....
~
~~540 *M"L-~~
..:.. . .... : . . . n. . ... _ u. .. . ._ -_ .- #-.- - . . . -
. . m__= -.... _.~._~ . ..'-. _._= - :. .- . : ::._
n_.- : n n.= t - .- -
_________a
~:n n: nz: .
-_.._..f_..__..._...__.._...~___..._.......-... . . ... .. ._. . _ .g........ ... ..... ... .
. ........._..,...i.._.
....). . . ... . . .....
. ... . .. . .................a....
. ... . .... .... ... s . ..... ...... . ._
......t
_.1.
.t........ .......... . ... ........_.....
t
.s...
....... .. .... . .. -....l......l...i......._...........l....._.l........... ..... ......... ... . . _.. .,.. .t............ ... . .... . . ...... .. ... - .
,.,...... ..........,t_..
80 100 120 0 20 0
'l'. Rate Core Power i
~~h. ,,2 4, . ,a..e~~
~~4. w ,~~
~~Fi_ ~. *h~~
~~Wa- cop *r ** 9 U U.N. MD DW a.~~
T;t0-1.00? 0? IRA ICN l 1
O.e 6 ') . a-1*-
-- - j
-Mis 2 e*
~~1 l~~
Figura TS.2.1-1 660 '
! ! l ! ! ! ! .
. .j . . ,. . , , . , , . . , ., ,
l 650 - . .. -- - - . .
i
- . . . j. . 4 , ..i.. .,g.. ., . i. . , ,
..;...i....p....<..
640 - . . . . .
630 --
..e . . . i. .
+ .
. i. . + ,.
..i.- +. ..i. .+.
. .+..-
LL :
o .. .. . . . . . . . .
y 620 -.' ~ ~ ~ .
..g . .;. ~3 , . .; . ;. . ;. . . . , . . . . ;. . . . ;. .
i+--
f 610 -' '
m . . . . , . . . .
s iii m
600 - . . -
.+ .
.i. - ... - , . . --
-, .. . - k E
e 590 -., , , , , . . , , .- . . . .-
1--
e . . . . . . . , , . . .
m fu 580 -. -- .- . - - - -
2385 psig O
..;. i
~
2235 570 -. - . - - .- . - - .-
560 -.
8
+ ,
1985 e 100% Flow (68.2 x 10 lb/hr) ' '
l1885l 550 - . - - -. - . - - . . . . . ,
l1785l 540 i ' ; ' ; ' ; ' - ' i ' - ' i 0 10 20 30 40 50 60 70 80 .
delta-T (Tn-Tc) F Reactor Core Safety Limits Figure TS.2.1-1 l 1
i l
l
, _ _ - . __~ _ - . . _ _ _ _
TS.3.4-1 RT! ?' Sl2Sl?1 l l
1 3.4 STEAM AND POWER CONVERSION SYSTEM l Applicability I l
Applies to the operating status of the steam and power conversion system.
Obiective l To specify minimum conditions of steam-relieving capacity and auxiliary feed- water supply necessary to assure the capability of removing decay heat from the reactor, and to limit the concentration of activity that might be released by steam relief to the atmosphere.
Specification A. Steam Generator Safety and Power Operated Relief Valves
1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied (except as specified in 3.4.A.2 below):
~~a. Ten steam generator safety valves shall be OPERABLE with lift settings of 1077, 1093, 1110, 1120 and 1131 psig [J4% except during testing.~~
~~b. Both steam generator power-operated relief valves for that reactor are OPERABLE.~~
2. During STARTUP OPERATION or POWER OPERATION, the following condition of inoperability may exist provided STARTUP OPERATION is discontinued until OPERABILITY is restored. If OPERABILITY is not restored within the time specified, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature below 350'F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,
~~a. One steam generator power-operated relief valve may be inoperable for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .~~
~~B. Auxiliary Feedwater System~~
1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied (except as specified in 3.4.B.2 below):
~~a. For single unit operation, the turbine-driven pump associated with that reactor plus one motor-driven pump are OPERABLE.~~
~~b. For two unit operation, all four auxiliary feedwater pumps are OPERABLE.~~
c. Valves and piping associated with the above components are OPERABLE i except that during STARTUP OPERATION necessary changes may be made in l motor-operated valve position. All such changes shall be under l direct administrative control. )
i Table TS.4.1-2A (Pagej(1{of])j RP' 106 f/21/93 MINIMUM FREOUENCIES FOR EOUIPMENT TESTS
, FSAR Sect.
E6uipmeht Test Freauency Reference i
1. Control Rod Assemblies Rod Drop Times of full length All rods during each refueling shutdown or 7 rods following each removal of the reactor vessel.
head; affected rods following maintenance ,
on or modification to the control rod drive system which could affect performance of those specified rods
~~2. Control Rod Assemblies Partial movement of all rods Every Quarter 7~~
~~3. Pressurizer Safety S:tpeintQ E Q UPER&BIX{Ih Per ASME Code,Section XI Inservice Testing. -~~
Valves *~a^EE6Vdiisiicetlwithi$lem(Inseidri{de Program .
ia m sp PEh TehN i g u e s$k sink h[N e n kI; g }
EEElIIIMfffffMGhthiedit
~~4. Main Steam Safety S: tp r intHV6HT)74iEMBtTttigiitiMid Per ASME Code,Section XI Inservice Testing -~~
Valves noordetwip
[IllWsFi$$iiiiis@*1$auseasi=jh{p"gProgram na M W e bhENkaYhb[ -
E
5. Reactor Cavity Water Level Prior to moving fuel assemblies or control e rods and at least once every day while the cavity is flooded. ;
M
~~6. Pressurizer PORV Functional Quarterly, unless the block valve has been ._ ."~~
~
~
Block Valves closed per Specification 3.1.A.2.c.(1),(b).2 L or 3.1.A.2.c.(1).(b).3. >
~~7. Pressurizer PORVs Functional. Every 18 months -~~
~~l, Table TS.4.1-2A IPig g fi bX 11 i MINIMUM FREQUENCIES FOR EQUIPMENT TESTS l~~
FSAR Sect.
EqGibiliiant Test Frecuency Reference
~~8. Deleted l~~
~~9. Primary System Leakage Evaluate Daily 4~~
~~10. Deleted~~
11. Turbine stop valves, Functional Turbine stop valves, governor valves and 10 governor valves, and intercept valves are to be tested at a intercept valves. frequency consistent with the methodology (Part of turbine presented in WCAP-ll525 "Probabilistic overspeed protection) Evaluation of Reduction in Turbine Valve test Frequency", and in accordance with the established NRC acceptance criteria for the I probability of a turbine missle ejection incident of 1.0x10-5 per year. In no case shall the turbine valve test interval exceed one year.
~~.-i~~
~~h 7~~
l d b
w
m
'_P c . f_ , /.
- 1_
_D _E*t.? 1. fi _c
- C,//., ,/ O_ *2_
1.
~~f. c.A_PF'_T'i_l~~
. _ T_T_ W
..F _. _ .j
.9
____b. __8. m_.5.m_. a.
m b .m M m a m &_ m_ , _ _ _a _ _ _
_ ___ _ _m_
m
..M
_m_e_m.
a
_=___ b..m_
/* m L 1
__ _I m _mmI a_ m _.._-1_1_
_ _ L_ _ ...6. I f _4 m_ J_
.. f __-_ -_ J f - 6. m_1_ j. .-
_ _. T_ k. m k. ._= 1_1__a_-m_
_ L_m- r-r-/
m e m-m & 1. .
w +m
.. s h. .m. J eka &La
__7 m_we.mJ--
_ _ 17_8 m. m D_ w a. _m i_ f a m e.- M. ._. ._1_. m_ _af*w* =.m w m e_- 4 m_ ._. _ = = _
. . _- . _ _ _ . _ _ _ _ _ _ _ _ PL a _iw. =
_ _ _ _ m_ _f _
.e. e_k. . ___I --_
w J a - #_ ,m. ._a e_.- a_ J__1_a_.m_w m6. - F .
c = f_ m F_,. . A.._. A .f. e.- #* - _- # e.-
. -e.-, a.m_ _. __-, A. __=_m_e_,.. 1_ ( __ _f e_.
w - e. w e- mh a11 La mwm mwmJ 'FL. f - ...a. L 11 J .m.,fkm s.e _I a 1_ _m &_ 4 m_ m _ . .
r--- - - - - - - - --
r- r--- - - - - - - -
r--- - - - - - - ----- ---
/1% mm 14mmL1m a f e n . .- m e- m m m m mw m A f m ,. .k. ..f m i a e. f /4s offm mf .L.
5r rr- ----- ---- - - - - - - -
r-------g - - - -

~~s 5i - - - - - - - - -- ----~~

s.s .I a.1. m e .I a_ m.
..a u m Ia a f i f 6-.m - m == m m .i 4- m .e m & m .s- a m .s- m e- w. . m 6. . . w m m a em J / dl \ &ka r-- ------/ --r-------u /- --- -- --------s -----
\i - - - -
ae
&_ _.1, m. _ _ m .e m a 1. . J m wa ..wwamm= 8PL. m es m m ., e- mk = 11 L_ m_ _m f_ .En w -_ m. a w w m_ m_ &_ _8 ._e m_ e. _I a.s.
-_ - -. 6. m.
r -- - - - - - - - - - - - - - - - - -
r--- ---- -
f a e.jse i_ _f _ f_ e_. . ..4 m. i_ _6_. f m_.
8PL
~~w . , e.~~
L_ e. m e.e _f
. ___J_ m.
em L_j. e &_ h. . _ A. gm. m_ __ w - e.- 4 m m_ _f.* . .m._ f &. e- a_ m..
_ _ . . . _ _a _ _ _
r---
mL.._m__i i L- .L 6Lm
_ _ _ I a_. 6_ _ J_
_ &_ a_ + L. . a_
. t_' a -- (
_ ____ m 4 m_ m. . , _ . . _ 17_f ._ _ D _., a_ m i__Jm _ _ _ _ . . .
e M. ._. 1 m_ _a w _ e_ __
_ e_ _a e_- f _ ,
_...J _ + L... . . . e. _a _f. j e ..
A.._. J_ f e.- f.**.-__-__fe.-&_*-.-
_ _ . 4 e-h. 4. m.
-_ e..._ _. . a l. m .F
_ _ _ _ . ._ e- L. . - _ ._...e..
m k. . _= 1_1_ .e_- _m _. &_k. . m. w f _ _ A_L,.
kn
/_b,. n_ w - e f a. m..
__ _ .. L_ m_ wm__m.-_
_ _ J_ ._. .. _ 6 f 1_ .. . __ _ M..._.._1_*m._ D._,.._.1_,e_._w..
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/* m -.m f a m i m m
B.2.1-1 Rr1 91 10/27/92 2.1 SAFETY LIMITf. REA NOP COFF P"_WTT1sndt6FC6M5ifstFNIEi w u Bases i To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturstion temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 and WRB-1 DNB correlations. The V-3 DNB correlation is used for Exxon fuel. The WRB-1 DNB correlation is used for Westinghouse fuel.
The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum valua of the DNB ratio, DNBR, during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 for the Exxon Nuclecr fuel luiisfthi?W3Mfisisti6n and to 1.17 for the Westinghouse fuel EsTugiths WRBilfes rheissioCThie re TiFaithifdIDNBK?lisIEspWEif Rs119]f6dhnsissd linE calcy$ akyschiM latiopsj s@ correspond These limits pu 6 $ $ $sto aM95%b ppyjj njfs M probability at afs95%
y y M Q3y @f~~
confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
i The safety limit curves of Figure TS.2.1-1 define the regions of acceptable '
I operation with respect to average temperatures, power, and pressurizer pressure. These boundaries of acceptable operations are limited by the thermal overpower limit (fuel melting), thermal overtemperature limit (cladding damage based on DNB considerations), and the locus of points where .
the steam generation safety valves open. These limits are used to set the l I
overpower and overtemperature AT trip setpoints.
l Ttis?K4fety?tisitBiditvss@;
Ts11 M insWrar4tega Qigd@Sj2[121@6mgisi3hipsbigingi; fig 6fithi[
E visss1 W i G sspuestafs E 650 2 1
ThisiliithiTassigs gesp6fitufsflimit g Thii?11mi gdsfinisythsip6ffiWd[6f thelsa fe tyflisi tf edrve s; fr6m;0?AT4of tbal firs tikne eilf6rsthe h2235jihd 3 238Mpsigicurve sM At3ths*RpressMasQtheithaperat6fejliEitiof]650@sf~my[s
B.2.1-2 om o, 2 n , ion pggtssstsPears*shistiftimiiq aa -s m n-u m ,- m c m -cons Bases continued Ti.d
~
tiid7Yee w RCS411di.tWThi.?I. ,u, cia.4N, f48t.W_ii.ii15,%. ifi1661EEsd?ffii yW M a,t ,it.-,. .n .j.minimum; mm .
~ .
~
1
. ..e..w ~m,wnex..h t <m w w.on~.,w.u wm w,,sen me.,in.
i 01g
.asmsw.>.ww D351531@iM35!!!MflUnst E.N~
na meas _uredy. .n
~~.NE.,,,n.-, }97v I~~
[litET.p.~d,Tr Ed d.7 ~ a e~1RP$ra.nN s
ew ftNh .~1b ihY re M ,a a.a.m
- m n.otru u u ti v -
er#Thi y
~~maa,sur,ne n oficor,s.,~~
won powergisy.-p roporti_manc.a.+.-,on,n onalgtescorer ,s.$ov;u finn >sk. y oniot isndil7857piti%unis $TtislisEdsist!y$$6i13ifssissins iu ed sfsch s~li~si. 1985i ilai~
h515MEsiMt$1j{hgesRigisssjipjinjspI}M~~~~
UDtiMs!uma!B!E11HGIIIM!MMEH3iMME As~aibati6nsd'bifera',7175"tr'ttW953FlilsitFNf' tit 4Wif@ir1*eW1 f tt053 critical beat flus correlation and'1,17,is thet BilBR' limit for Westinghouse fuel using,ttie WRS-1, critical heat flus'correlatissi"*ths locus of points past'the first knee at*all pressures 1 represents,the
' thermal-hydraulic conditionis above which,the tiot/ anaal h bas' a'DNBRJ1ess than^the-limit.~TtC oonditions are evaluated using approved DNB inschodology'( The assumptions used in' the calculation include, 'a hypsis flow of 64,s an Fa' greater ttian 1.75, and s; rod bow penalty of 2.64.' Thd very shallow knee at' full power AT occurs because the Fd (hot channel power) is allowed to increase' for core po;wer lass,tharLRATED,1MRMAi; _
POWER as described ~iti TS 3.10,B.1!
91Hgg@li$iig@ulity311}{ypp0DQij@fsgi]{ppl@' p My 7h r6 4
Thia"flimiWiFitypisil.~
~~ x .
approached ~istilowerepowers.,s.
.r lyn. ~6fEth_ ha.re:MtTwftTgo+rg6$0*W sis,ai,tWsiWi.v#1V.s?bs~cussT,istmore1Eis4s.
-- n -w --
Hsis112 i' limiting $$$who d icMATsshsihiidd$tsestins 3RESEss1NinEid$sIMidifibsd kboussard$sifd6sedissisdapphiud?DetBliisnth3ds15djjjifidijnsiWUs~~
asc'ewsin. d4b - s 11cyf,.fi.=$..
h.aeeflu s
x ncorr yy~.14ttis. w.tion.
Wing,Moty,,the(shnEn,s1,Es_ife,%~
. s Thet ~ w xt.u ,erchannetsex
. n tt,avalte.,tuRufisfait n
i fpj[tje[y[ 6 $ $ ilf51$n M [M Q fjrl{tislRRB11j j pof M h t,ysitettjaist
~ ~n ope ration ~ abovs" thF safety'lisit'~chryss74f'Fighis~'TS ~ 2."1f17 ti~ riot"acssptabis'[']
At each pressure the safety limit' curve is the most restrictive combination of, the four limits discussed "above. Tho' area 'of acceptable operation,below~
the safety' limit curves 'is bounded by the OTAT trip , the;0 PAT trip, 'and ttis]
locus _o ' f points where' tho'st'eam generator ,(main stosa), aafety valves,open!
the AT trips are, set conservatively with respect to'the safety limit curves to protect the core fros, exceeding the' safety limits. The: locus' of, points at which the, steam generator sofety valves open defines the thermodynamic limit of temperature conditions in the RCS based on theimaximum pressure in the steam geneirators. For this calculation, it'is assumed that the pressure in the steam generator is 1195.psig which;is,1104_of, design pressure, iIt isj t
.-w,,%.. ---- - , , ,,-.r, , - ,,
4 B.2.1-3 A. Reactor Core Safety Limits l
Bases continued J "th.-Q.a.-
~....ci,e ...x:... .....;...w
_.; x --
.._t?x..hej. t. . . ,.~pt_es,atsgs*.f:
. . ; p.;yy , a:. .c n-gid.
a,.~. a iM
.~,,7
,w&.-
$sese, ditilg?A.,ld%w..y?#f.d4.-
w n v w ,,,ee i. press. ure$,s,.,,.te.m. ~
~~,;g. tla;sg- .<,,1.1. ; -~~
-u- .,w.
.x ;
. s . A sn $f5fM sfitidijiii
$N .: hh $e - ~e .i , n.;$ . . $.
I._ 3p
,. tad. tsanyw$s jdsysisid,~$;1,1_stE_5f,ittE_IAT$t_ztN.
. - a n-- is,s_di~k_tlM.....
np_e. i .g.-n, i -~~--
h*neratorZ.~
-_m n. n,,~Swee2 ive m As'"isiTesiisp1'751Ptll471'imitF#6Ktti&T2f3fp41DiijNNifafilF"plstisd 4 Jif figiiM n;t'11:alongivich the AT trips and the teams of'petats innere;ttie. steam
^
generator safetyNaisee ' spear 3his iplet',dammmatzstes;tGt the AT trips ~ssid the ' steam generate ( safety estwee' de preteet' theireacter 'from'esseeding the'
'afety s limitsr' Note'.' housvari that the 'OT&T, trip ,teous',5a that,' plot is for
's teady stets'eamsttlagins'and that the leeus wt11^4refia' responsete;tho5rstid t
he ubich the AT .is1.1 oar.eas.i.ng. _ita.."ad_dit.~i.o_ni 'f.(A. f..r.a nar.e,~a,s..ing i _ v s._i.l..i.._ais._e _lo.wer th O_ TAT.t.
- - RIP leou_s.64 Aa -~
Thui.sif.s_tfEll.mimitIWervesYsfeD%stisdMthT.&TDni@.m.~isIffEEiiffsi.e
%u mm m._ m m -
tt mme.y f, sfEtti.s,?f611F.m ee G.
m mmm i,.iiiis~.
w.v . .[.e= ~ 8;+: s <c p f, .icult4tsiplotfiltt.a 2 U.e).r.+ rotection h,.w.;;;p&
r -
thovremotorB
~nm#tnek Titr,i,p;;;;w n.setpem.~>la:,tsiwhicO.:.< .mwne wn;;wya.w ,.,.+.,;z sys._
tea _ra_ctu.a_ll~ye nc.~a_lcul._ates s.
ess di. w.
eSnw;c. ass;;;.< ;gy e a;.;e;;&wcc;s..,a~s.e_d, n ype.rcent9.
--- powers ion._;.th.w~
":: th: :::rt:r;:rstur: li=it, the fell:uing f:ur liriting crit:ri cre used4 1_ . u __ _ _ i_ . e ._. . _ _ - . _ . . _ . . _ . . _ . _ .
- r. e. n . c. r,a _ _ _ _ , . . . _ _ - . _ _ . _ . _ . . _,4._
__ _s.
~~2. V :::1 : it t:r;:retur: '~~
~~tur:ti:: t r;:r:tur: 'encur:: ; uer ^T).~~
1
~~u. .nu. . n .o. s 1_ . o_ ig_r._. _1_~~
a.____ 1_._ e ._. _r ._, s ,
g_
~~i. u.._,. _t..___.._._,_~~
_ ._ .._.._.s.e._...
....3- . j . s. .c . i,4_._._.._.
3_ .
-. c.u. . .e__.._s____<-__s.
S:: fir:t tu: criteri: re: ult ir : cingle lirit er 7::::1 exit t ; ::-
._._.w.... v. ...._t.._ 3_ c_ o_ c_.,._e,- _. 2_ ,.o.__o ,__ c .._4,_ . _ . _ . _ __, ._t..__ ___,__._ _ . . . . . _ , _
__.t.._,.. .. .t_ __._ . . __..., __......_2 . . . . ...t_i_., t_,_..
.r> -- - - - - ' ---- - - - - - --
7-- -- - - - - - - - - - - - -----

~~r/ ----"~~

___._.,_.._3_. _c o_ s_ .___2 _.._
, i. .
.______._.4._.._3_,... r_ _ . .t.._ ,,,c 7__ __4,_ _ . 2_ o s o_ _c 7___. __ _ _ _. .. . _ , _ _ __ . ____ __
peig curfte, th :::1:nt ever:g: t r;:retur: et th: ::r: : it 1: :7z.-1
__2 ,
.__ c_ e n_ . c.t_ __ _3_,___
_ _.____ _ . . _ . _ i _ ._. _ $_ _ _ c c. i . _ __ . _._.2.. . ___,__ _ _ _ __ _ . _ . _ . _ , . .
~~f,'L..._~~
. . n. . .f w J.-
. _..J_
_f _ ._. w . t. .
_ _. .. .._I._._#_. -..._....i....e_J_ ._._f,_
_ __J.._J
-. h_u. .n.
.a_
-. . . t...
ch. :I .eur */"[:: I : : 1I I Ch I . I [:u I _:'f _e.
5: cret ef :sf: peratier in heleu th :: curz::.
i I
. 1 E5Nb EggtssEEsi3NF#ifsWYTEit!if Bases continued 1
The-plent rditier required te viel:t the lirit ir the'1:rer p:rer i r r;: ::: pre:1rord by th: ::1f-::turt2d : fety relre er th trer
rreretere. The high

t nerir:1 ::ttir; ef t' :trer ;:rerrter erfety valver 1: 1129 prig ( :tur: tier t:rp retur: 550*F). ^t ::re p:rer t'-
~~i differere: heturer pristry :: lert med ::: rdery :: lert 1: :r cro et full perer it 1: 50*F The reerter : rditirr et rhich trer ; rer:ter~~
~~fety relver per i: cherr :: : '--hed lire er Fi;ur TS.2.1-1.~~
Except for special tests, POWER OPERATION with only one loop or with natural circulation is not allowed. Safety limits for such special .
l tests will be determined as a part of the test procedure.
The curves are conservative for the following nuclear hot channel factors: l F
~~-F RTP~~
[1 + PFDH(1-P)) ; and F" - FATP a a o o :
where:
-FATP is the Fg limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.
-FRTP is the Fu limit at RATED THERMAL POWER specified in the CORE a
OPERATING LIMITS REPORT.
- PFDH is the Power Factor Multiplier for F# specified in the CORE OF . TING LIMITS REPORT Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.
This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified in the CORE OPERATING LIMITS REPORT assure that the DNB ration is always greater at part power than at full power.
The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for' Vestinghouse fuel. <
B . 2. E53-4 REV 91 10/27/93 trggnsaatwc66nw Wsurwas s~ersu ramla re
2. 2 ?? Fe'N LIMIT "EACT^" C^^UJ?T SYSTEM " ESSURE Eases The reactor coolant system (Reference 1) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmos-phere. In the event of a fuel cladding failure the reactor coolant system is the primary barrier against the release of fission products.
By establishing a mystem pressure limit, the continued integrity of the reactor coolant system is assured. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III is 110% of design pressure.
The maximum transient pressurs allowable in the reactor coolant system piping, valves and fittings under USAS Section B31.1 is 120% of design pressure. Thus, the safety limit of 2735 psig (110% of design pressure) has been established (Reference 2).
The nominal settings of the power-operated relief valves, the reactor ;
high pressure trip and the safety valves have been established to assure l that the pressure never reaches the reactor coolant system pressure safety l limit.
1 I
In addition, the reactor coolant system safety valves (Reference 3) are sized to prevent system pressure from exceeding the design pressure by more than 10 percent (2735 psig) in accordance with Section III of the ASME Boiler and Pressure Vessel Code, assuming complete loss of load without a direct reactor trip or any other control, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valves settings.
As an assurance of system integrity, the reactor coolant system was hydrotested at 3107 psig prior to initial operation (Reference 4).
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B.3.1-2 REV 106 5/21/92 3.1 REACTOR COOLANT SYSTEM Bases continued A. Operational Components (continued)
Reactor coolant pump start is restricted to RCS conditions where there is pressurizer level indication or low differential temperature across the SG tubes to reduce the probability of positive pressure surges causing j overpressurization.
l The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point.
PMThssWWE1W34TsNi*s364TddsdTOPERABLEinUF34@iiiigitiK11@iistiiniEWsO4 Big?!S$
h k g g nnainal M Q d f M p;;setpoin [ Below 350*F and 450 psig in the kb!i b reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F. The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load (Reference 1).
The requirement that two groups of pressurizer heaters be OPERABLE provides assurance that at least one group will be available during a loss of offsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions.
The pressurizer power operated relief valves (PORVs) operate to relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve .
become inoperable. The PORVs are pneumatic valves operated by instru- !
ment air. They fail closed on loss of air or loss of power to their DC solenoid valves. The PORV block valves are motor operated valves supplied I by the 480 volt safeguards buses.
The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions: )
I
~~a. Manual control of PORVs to control reactor coolant pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown.~~
~~B.3.1-3 RE" 105 5/2L/S3 3.1 REACTOR COOIANT SYSTEM Bases continued A. Operational Components (continued)~~
~~b. Maintaining the integrity of the reactor coolant pressure boundary.~~
~~This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.~~
c. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item b. above).
~~d. Manual control of a block valve to isolate a stuck-open PORV.~~
The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS temperature is less than 310*F*.
The PORV control switches are three position switches, Open-Auto-Close. A PORV is placed in manual control by placing its control switch in the Closed position.
The minimum pressurization temperature (310*F *) is determined from Figure TS.3.1-1 and is the temperature equivalent to the RCS safety relief valve setpoint pressure. The RCS safety valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low temperature operational transients. Inadvertent pressurization of the RCS at temperatures below 310'F* could result in the limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded. Thus the low temperature overpressure protection system, which is designed to prevent pressurizing the RCS above the pressure limits specified in Figures TS.3.1-1 and TS.3.1-2, is enabled at 310*F*. Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limits of Figures TS.3.1-1 and TS.3.1-2 from being exceeded.
The setpoint for the low temperature overpressure protection system is derived by analysis which models the performance of the low temperaP.ure overpressure protection system assuming various mass input and heat input transients. The low temperature overpressure protection system setpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 l and TS.3.1-2) are revised.
The 3 square inch RCS vent opening is based on the 2.956 square inch cross l sectional flow area of a pressurizer PORV. Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.
~~Valid until 20 EFPY C~~
B.3.4-1 REV 9' 6/26/91 3.4 STEAM AND POWER CONVERSION SYSTEMS Bases A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.
The ten steam generator safety valves have a total combined rated capability of 7,745,000 lbs/hr. The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten steam generator safety valves will be able to relieve the total ateam flow if neeeasary (Referenee 1).2iissiiiFiFsliisE7& ifs 74siiis18sEsdEOPERABLEM 75st*E3F6fittnir1WiiijRiistf15dTsit5615@Ys115GingMNtN$^}jIl%~ nt'6ENNiiii@s]d ~ - ~ ~~~
ILMssisis1Mnt=Ess1M11GEMMinistM In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be assured by availability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the steam generator safety valves.
One auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedwater pump for each reactor can be made available to the other reactor. During STARTUP OPERATIONS, the Auxiliary Feedwater motor-operated injection valves maybe less than full open as necessary to faciliate plant startup.
The minimum amount of water specified for the condensate storage tanks is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shutdown. Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling '
water system.
The two steam generator power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a steam generator tube rupture event (Reference 3) and following a high energy line rupture outside containment (Reference 2). The ,
steam generator power operated relief valves are provided with manual upstream J block valves to permit testing at power and to provide a means of isolation.
In order to assure timely response to a steam generator tube rupture event, a steam generator power operated relief valve is considered operable when it is capable of being remotely operated and when its associated block valve is open. i
- Isolation dampers are required in ventilation ducts that penetrate those rooms containing equipment needed for a high energy line rupture outside containment. j I
l
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I The limitations on secondary system specific activity ensure that the resultant ,
! off-site radiatian dose will be limited to a small fraction of 10 CFR Part 100 l limits in the aveat of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpa primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the i assumptions used in the accident analyses. l Feferences i
~~1. USAR ::eci.xen 11.9.4 !~~
~~2. USAR, Appendix I~~
~~3. US/.2, Section 14.5.4 l~~
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.n LICENSE AMENDMENT REQUEST DATED May 4,1995 l Pressurizar Safety Valves and Main Steam Safety Valves Lift .[
i Settina Tolerance channe and Safety 13=it Curves Channes i i
i EXHIBIT C i i
t Appendix A, Technical Specification Pages Revised Pages 3 TS-i i TS-viii l TS-x TS-xiii I TS.2.1-1 j TS.2.3-2 i TS.2.3-3 Figure TS.2.1-1 !
TS.3.4-1 :
Table TS.4.1-2A (Page 1 of 2) :
Table TS.4.1-2A (Page 2 of 2) gj B.2.1-1 i B.2.1-2 i B.2.1-3 l B.2.1-4 i B.2.1-5 ;
Figure B.2.1-1 ;
B.2.2-1 j B.3.1-2 l B.3.1-3 'l B.3.4-1 j B.3.4-2 ,
6 I
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TS-1 4
TECHNICAL SPECIFICATIONS IbBLE OF CONTENTS If_f A '72E TITLE PAGE ,
1.0 DEFINITIONS- TS.1-1 l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING TS.2.1-1 ,
2.1 Safety Limits TS.2.1-1 :
2.2 Safety Limit Violations TS.2.1-1 2.3 Limiting Safety System Settings, Protective '
Instrumentation TS.2.3-1 i
A. Protective Instrtraentation Settings for Reactor >
Trip TS.2.3-1
~~3. Protective Instrumentation Settings for Reactor ,~~
Trip Interlocks TS.2.3-4 C. Control Rod Withdrawal Stops TS.2.3-4 ;
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TS-viii ,
TABLE OF CONTENTS (Continued) j TS SECTION TITLE PAGE .
t 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1 j 6.1 Organization TS 6.1-1 6.2 Review and Audit TS.6.2-1 A. Safety Audit Committee (SAC) TS.6.2-1
~~1. Membership TS.6.2-1~~
~~2. Qualifications TS.6.2 1~~
~~3. Meeting Frequency TS.6.2-2~~
~~4. Quorum TS.6.2-2~~
~~5. Responsibilities TS.6.2-2 r~~
~~6. Audit TS.6.2-3~~
~~7. Authority TS.6.2-4 [~~
~~8. Records TS.6.2-4 ;~~
~~9. Procedures TS.6.2-4 t B. Operations Committee (OC) TS.6.2-5~~
~~1. Membership TS.6.2-5 '~~
~~2. Meeting Frequency TS.6.2-5~~
~~3. Quonus TS.6.2-5 .~~
~~4. Responsibilities TS.6.2-5 !~~
~~5. Authority TS.6.2-6 !~~
~~6. Records TS.6.2-6 l~~
~~7. Procedures TS.6.2-6 ~~~
C. Maintenance Procedures TS.6.2-7 6.3 Special Inspections and Audits TS.6.3-1 6.4 Deleted )
6.5 Plant Operating Procedures TS.6.5 1 ,
A. Plant Operations TS.6.5-1 l B. Radiological TS.6.5-1 ;
C. Maintenance and Test TS.6.5-3 D. Process Control Program (PCP) TS.6.5-3 !
E. Offsite Dose Calculation Manual (ODCM) TS.6.5-4 F. Securtiy- TS.6.5-4 C. Temporary Changes to Procedures TS.6.5-4 6.6 Plant Operating Records TS.6.6-1 A. Records-Retained for Five Years TS.6.6-1 B. Records Retained for the Life of the Plant TS.6.6-1
L I
TS-x TABLE OF CONTENTS (continued) ,
l TS BASES SECTION TITLE PACE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM '
SETTINGS 2.1 Safety Limits B.2.1-1 A. Reactor Core Safety Limits B.2.1-1 B. Reactor Coolant System Pressure Safety LLmits B.2.1-5 2.2 Safety Limit Violations B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION ,
3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4-1 3.5 Inscrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 B. Caseous Effluents B.3.9-2 i I
C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 E. 6 F. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 i G. Contre' Rod Operability Limitations B.3.10-9 l H. Rod Drop Time B.3.10-10 l I. Monitor Inoperability Requirements B.3.10-10 i J. DNB Parameters B.3.10-10 ;
3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Fire Detection and Protection Systems B.3.14-1 3.15 Event Monitoring Instrumentation B.3.15-1
TS-xtit APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Reactor Core Safety Limits 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >l.0 uCi/ gram DOSE EQUIVALENT I-131 3:8-1 Spent Fuel Pool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Caseous Effluents 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Region Minimum Burnup Requirements B.2.1-1 Origin of Safety Limit Curves at 2235 psig with delta-T Trips and I Locus of Reactor Conditions at which SG Safety Valves Open l
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TS.2.1-1 2.0 FAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMITS A. Reactor Core Safety Limits In MODES 1 and 2, tThe combination of thermal power (measured in aT), pressurizer pressure, and the highest reactor coolant system loop average temperature shall not exceed the limits shown in Figure TS.2.1-1.
B. Reactor Coolant System Pressure Safety Limit In MODES 1, 2, 3, 4, and 5, the reactor coolant system pressure shall not exceed 2735 psig.
2.2 SAFETY LIMIT VIOLATIONS
)
A. If SAFETY LIMIT 2.1.A. is violated, restore compliance and be in MODE 3 l
within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
B. If SAFETY LIMIT 2.1.B. is violated:
l
~~1. In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . I~~
~~2. In MODE 3, 4, or 5, restore compliance within 5 minutes.~~
C. If a SAFETY LIMIT is violated, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notify the NRC Opcretions Center in accordance with 10CFR50.72.
D. If a SAFETY LIMIT is violated, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notify the Vice President Nuclear Generation, and the Chairman of the Safety Audit Committee or their designated alternates.
E. If a SAFETY LIMIT is violated, within 30 days a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the Vice President Nuclear Generation and the Safety Audit Committee.
F. If a SAFETY LIMIT is violated, operation of the unit shall not be resumed until authorized by the NRC.
I
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TS.2.3-2 2.3.A.2.d Cont.
and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chamber, with gains to be selected based on measured instrument response during plant startup tests, such that where qs and qe are the percent !
power in the top and bottom halves of the core, respectively, and qs + gn is total core power in percent of rated _ power:
~~1. for qt - qu within -12t and +96, f (AI) - 0, and l~~
~~2. for each percent that the magnitude of q - q. exceeds~~
~~+94 the AT trip set point shall be automatically reduced by an equivalent of 2.5 percent of RATED THERMAL POWER.~~
~~3. for each percent that the magnitude of qs - qm exceeds~~
~~-124, the T trip set point shall be automatically reduced by an equivalent of 1.5 percent of RATED THERMAL POWER.~~
~~e. Overpower a T K 33 t sT~~
'f p s Aro tr4 ~ ~ K6(T-I' 3 ~ f (AI)3 1+ts3 where AT, - Indicated AT at RATED THERMAL POWER T - Average temperature. *F >
T' -~ 567.3*F ;
K. s 1.10 ,
K3 - 0.0275 for increasing T: 0 for decreasing T l K, - 0.002 for T > T', O for T < T' ]
t3 - 10 sec f(AI) - as defined in d. above
~~f. Low reactor coolant flow per loop - t90% of normal indicated loop flow as measured at loop elbow tap.~~
l
i; TS.2.3-3 -
2.3.A.2.g. Open reactor coolant pump motor breaker,
~~1. Reactor coolant pump bus undervoltage - ,~~
~~E75% of normal voltage.~~
~~2. Reactor coolant pump bus-underfrequency -~~
~~E58.2 Hz~~
~~h. Power range neutron flux rate.~~
~~1. Positive rate - s15% of RATED THERMAL POWER with a time constant E2 seconds ,~~
~~2. Negative rate - 57% of RATED THERMAL POWER with a time constant 22 seconds~~
~~L 5~~
~~3. Other reactor trips~~
~~a. High pressurizer water level - 590% of narrow range. instrument span,~~
~~b. Low-low steam generator water level - E5% of narrow range instrument span.~~
~~c. Turbine Generator trip ,~~
~~1. Turbine stop valve indicators - closed~~
~~2. Low auto stop oil pressure - 245 psig~~
~~d. Safety injection - See Specification 3.5 E~~
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E e 590 -. . . . . . . . , , . . . -
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en E 580 -- - - - -
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> ..j. ., ; ..;.. ( [ ..<....).. ..(.. . .i . , .(...-
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2235 570 -- - . - . - . . . . .-
i 560 -- . .i - - -i - -i ' -
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1
. l1885 l 550 -
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. . . . J 540 i i i ' ; ' ; ' i ' : ' ; ' .
0 10 20 30 40 50 60 70 80 delta-T (Tn Jc) T !
Reactor Core Safety Limits Figure TS.2.1-1
. _,_ _ __.m . _ . . . _ . . _. _ _ _ _ . . . _ . __
I
.. 1 TS.3.4-1 E
i 3.4 STEAM AND POWER CONVERSION F O( !
t i
Acolicability i
Applies to the operating status of the steam and power conversion system. {
Obiective ,
To specify minimum conditions of steam-relieving capacity and auxiliary feed- water !
supply necessary to assure the capability of removing decay heat from the reactor, ;
and to limit the concent ation of activity that might be released ;
by steam relief to the atmosphere.
Specification i
A. Steam Generator Safety and Power Operated Relief Valves ,
1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following l conditions are satisfied (except as specifie6 in 3.4.A.2 below): ,
~~a. Ten steam generator safety valves shall be OPERABLE with lift settings of 1077, 1093, 1110, 1120 and 1131 psig i 36 except during testing. l~~
~~b. Both steam generator power-operated relief valves for that reactor are OPERABLE. ;~~
~~2. During STARTUP OPERATION or POWER OPERATION, the following condition of inoperability may exist provided STARTUP OPERATION is discontinued until ;~~
OPERABILITY is restored. If OPERABILITY is not restored within the time ;
specified, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature below 350*F within the ;
following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . l
~~a. One steam generator power-operated relief valve may be inoperable ,~~
for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
B. Auxiliary Feedwater System ,
L
1. A reactor shall not be made or maintained critical nor shall. reactor coolant system average temperature exceed 350*F unless the followin5 conditions are satisfied (except as specified in 3.4.B.2 below):
~~a. For single unit operation, the turbine-driven pump associated with that reactor plus one morcr-driven pump are OPERABLE.~~
~~b. For two-unit operation, alt four auxiliary feedwater pumps are OPERABLE.~~
~~c. Valves and piping associated with the above components are OPERABLE l except that during STARTUP OPERATION necessary changes may be made in '~~
motor-operated valve position. All such changes shall be under direct administrative control.
l
l Table TS.4.1-2A (Page 1 of 2)
MINIMUM FREOUENCIES WR EQUIPMENT TESTS FSAR Sect.
Eculoment Test Freauency Reference l
1. Control Rod Assemblies Rod Drop Times of full length All rods during each refueling shutdown or 7 rods following each removal of the reactor vessel head; affected rods following maintenance on or modification to the control rod drive system which could affect performance of those specified rods
~~2. Control Rod Assemblies Partial movement of all rods Every Quarter 7~~
~~3. Pressurizer Safety Verify OPERABLE in accordance Per ASME Code,Section XI Inservice Testing -~~
~~Valves with the Inservice Testing Program Program ( 34). Following testing, lift settings shall be within 1%~~
~~4. Main Steam Safety Verify each required lift Per ASME Code,Section XI Inservice Testing -~~
Valves setpoint in accordance with Program the Inservice Testing Program (t 3%). Following testing, lift settings shall be within 1%
QY
~~5. Reactor Cavity Water Level Prior to moving fuel assemblies or control $$~~
~~rods and at least once every day while the cavity is flooded. "d ab~~
~~6. Pressurizer PORV Functional Quarterly, unless the block valve has been _~~
~~sa r' Block Valves closed per Specification 3.1.A.2.c.(1).(b).2 3 or 3.1.A.2.c.(1).(b).3. >~~
~~7. Pressurizer PORVs Functional Every 18 months -~~
t
-a
. v e Table TS.4.1-2A (Page 2 of 2)
MINIMUM FREOUENCIES FOR EOUIPMENT TESTS FSAR Sect..
Eauipment Test Frecuency Reference l
~~8. Deleted~~
~~9. Primary System 1.eakage Evaluate Daily 4~~
~~10. Deleted~~
11. Turbine stop valves, Functional Turbine stop valves, governor valves and 10 governor valves, and intercept valves are to be tested at a intercept valves. . frequency consistent with the methodology (Part of turbine presented in WCAP-ll525 "Probabilistic overspeed protection) Evaluation of Reduction in Turbine Valve test-Frequency", and in accordance with the established NRC acceptance criteria for the probability of a turbine missle ejection incident of 1.0x10'5 per year. In no. case shall the turbine valve test. interval exceed one year.
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_ _ _ _ _ _ _ - , _ - - - _ . . - _ _ _ _ _ . _ . . . . . - . . ~ . . _ . _ _ . . - . - . .
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B.2.1-1 2.1 SAFETY LIMITS l A. Reactor Core Safety Limits Bases r To maintain the integr'ity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat :'
transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter i during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB !
through the W-3 and WRB-1 DNB correlations. The W-3 DNB correlation is used ;
for Exxon fuel. The WRB-1 DNB correlation is used for Westinghouse fuel. .
The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the i heat flux that would cause DNB at a particular core location to the local !
heat flux, is indicative of the margin to DNB. The minimum value of the DNB !
ratio, DNBR, during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 for the Exxon fuel using the W- l 3 correlation and to 1.17 for the Westinghouse fuel using the WRB-1 correlation. There is a third DNBR limit specifically for the steam line break accident but it does not apply to the safety limit curve calculations.
These limits correspond to a 954 probability at a 954 confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The safety limit curves of Figure TS.2.1-1 define the regions of acceptable operation with respect to average temperatures, power, and pressurizer 1 pressure. These boundaries of acceptable operations are limited by the i thermal overpower limit (fuel melting), thermal overtemperature limit ,
(cladding damage based on DNB considerations), and the locus of points where the steam generation safety valves open. These limits are used to set the overpower and overtemperature AT trip setpoints.
The safety limit curves of Figure TS.2.1-1 comprise the most limiting of the i following four criteria: )
l
1) Vessel Exit Temperature < 650*F This is the design temperature limit. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 2235 and 1 2385 psig curves. At these pressures, the temperature limit of 650' is more l
~~B.2.1-2 A. Reactor Core Safety Limits Bases continued limiting than the T,,s limit. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.~~
2) Vessel Exit Temperature < T.,s This limit ensures that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated water which ensures that the AT measured by instrumentation used by the RPS as a measure of core thermal power is proportional to core power. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 1985, 1885 and 1785 psig curves. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.
3) Minimum DNBR > 1.3 or 1.17 whichever is applicable As mentioned before, 1.3 is the DNBR limit for Exxon fuel using the W-3 critical heat flux correlation and 1.17 is the DNBR limit for Westinghouse fuel using the URB-1 critical heat flux correlation. The locus of points past the first knee at all pressures represents the thermal-hydraulic conditions above which the hot channel has a DNBR less than the limit. The conditions are evaluated using approved DNB methodology. The assumptions used in the calculation include a bypass flow of 64, an Fan greater than 1.75, and a rod bow penalty of 2.6%. The ,
~~very shallow knee at full power AT occurs because the Fas (hot channel power) is allowed to increase for core power less than RATED THERMAL POWER as described in TS 3.10.B.1.~~
4) Hot Channel Exit Quality < 15% or 30% whichever is applicable This limit is typically not the most restrictive because it is generally approached at lower powers where the T,nt < T,,e or 650*F is more limiting. However, it is considered when the DNB calculations described above are performed using approved DNB methodology. This limit is determined by the range of the channel exit quality for the critical heat flux correlations. The maximum channel exit quality limit is 15%
for the W-3 correlation and 30% for the WRB-1 correlation.
Operation above the safety limit curves of Figure TS 2.1-1 is not acceptable.
At each pressure the safety limit curve is the most restrictive combination of the four limits discussed above. The area of acceptable operation below the safety limit curves is bounded by the OTAT trip, the OPAT trip, and the locus of points where the steam generator (main steam) safety valves open.
The AT trips are set conservatively with respect to the safety limit curves j to protect the core from, exceeding the safety limits. The locus of points at which the steam generator safety valves open defines the thermodynamic limit l of temperature conditions in the RCS based on the maximum pressure in the i steam generators. For this calculation, it is assumed that the pressure in ,
the steam generator is 1195 psig which is 110% of design pressure. It is l l
- 1 l
1 B.2.1-3 A. Reactor Core Safety Limits Hagga continued required that the steam generator safety valves protect the pressure from exceeding 110% of design pressure so using 1195 psig in the calculations is conservative. Thus, the reactor is protected from violating the safety limits by the physical limit of the AT trips and the opening of the steam generator safety valves.
As an example, all the limits for the 2235 psig curve are plotted in Figure B.2.1-1 along with the AT trips and the locus of points where the steam generator safety valves open. This plot demonstrates that the AT trips and the steam generator safety valves do protect the reactor from exceeding the safety limits. Note, however, that the OTAT trip locus on that plot is for steady state conditions and that the locus will drop in response to the rate at which the AT is increasing. In addition, f(AI) increasing will also lower the OTAT trip locus.
The safety limit curves are plotted with AT on the x-axis for the following two reasons: 1.) the full power AT is different at different temperatures and pressures because water properties are nonlinear. This makes it difficult to plot the curves at each pressure using the same scale for the percent power axis. 2.) the AT trip setpoints which the reactor protection system actually calculates is based on the AT, not the percent power.
Except for special tests, POWER OPERATION with only one loop or with natural circulation is not allowed. Safety limits for such special tests will be determined as a part of the test procedure.
The curves are conservative for the following nuclear hot channel factors:
F" -F RTP (1 + PFDH(1-P)] ; and F#-FRTP An an o o where:
-F RTP is the Fn limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.
-FRTP is the F s6 limit at RATED THERMAL POWER specified in the CORE an OPERATING LIMITS REPORT.
- PFDH is the Power Factor Multiplier for F" specified in the CORE OPERATING LIMITS REPORT i Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.
l 1
B.2.1-4 l A. Reactor Core Safety Limits Bases continued This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified in the CORE OPERATING LIMITS REPORT assure that the DNB ration is always greater at part power than at full power.
The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel.
I l .
1 1
i l
4 M
B.2.1-5 l B. Reactor Coolant System Pressure Safety Limit Bases The reactor coolant system (Reference 1) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmos-phere. In the event of a fuel cladding failure the reactor coolant system is the primary barrier against the release of fission products.
By establishing a system pressure limit, the continued integrity of the reactor coolant system is assured. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III is 110% of design pressure.
The maxinum transient pressure allowable in the reactor coolant system piping, valves and fittings under USAS Section B31.1 is 120% of design pressure. Thus, the safety limit of 2735 psig (110% of design pressure) has been established (Reference 2).
The nominal settings of the power-operated relief valves, the reactor high pressure trip and the safety valves have been established to assure that the pressure never reaches the reactor coolant system pressure safety limit.
In addition, the reactor coolant system safety valves (Reference 3) are sized to prevent system pressure from exceeding the design pressure by more than 10 percent (2735 psig) in accordance with Section III of the ASME Boiler and Pressure Vessel Code, assuming complete loss of load without a direct reactor trip or any other control, except that the safety valves on the secondary plant are assumed to open when the steam l pressure reaches the secondary plant safety valves settings.
1 As an assurance of system integrity, the reactor coolant system was hydrotested at 3107 psig prior to initial operation (Reference 4).
t l
s
4
. . . i Figura B.2.1-1 ,
i i
l ! ! ! ! I i a . I.
650 .+ . .+ - - -
-.+..- ..-
... . ....p. , , . .j . ,
. .g . ,
. .} . . .;.. , ,
j 640 -. - \
- . . . Exit Temp h.
7 Limit 650 F . . . .
630 v .+ ..t. . .; . n. . .v .r. .v c...-
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o Q 620 -. . ~ .- - , . ..-
U h ... . .j . ..j.. ,
..i. ,
,,y., .
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4 f 610 - - - - - -
e . Locus of Reactor . .. . ,
f Ei Conditions at which .
y 600 -
the SG Safety > ' '
i-
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Valves Open E
e 590 -, ,
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f'e 580 -. - . . - - - - - . . -
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.3 < , '
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570 --- . - - - - . .
Limit i 4
Region of . . OTDT l 560
.+ ' '
t .t . Operation
't "j" Trip 550 - ' ' ' ' ' ' '
OPDT
_ .; . .i. 4 , , , . .; . , . .; . , . Trip 540 - ' i i ; ' i ' ; ' ; i ; i I O 10 20 30 40 50 60 70 80 delta-T (Tn-Tc) F Origin of Safety Limit Curves at 2235 psig with delta-T Trips l 4
and Locus of Reactor Conditions at !
which the SG Safety Valves Open Figure B.2.1-1 1
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B.2.2 1 2.2 SAFETY LIMIT VIOLATION 1 Bases If the reactor core SAFETY LIMIT 2.1.A is violated, the requirement to go to ;
MODE 3 places the unit in a MODE in which this SAFETY LIMIT is not ;
applicable.
The allowed completion time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SAFETY LIMIT is not applicable, and reduces the probability of fuel damage.
If the Reactor Coolant System pressure SAFETY LIMIT 2.1.B is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , t
Exceeding the Reactor Coolant System pressure SAFETY LIMIT may cause immediate Reactor Coolant System failure and create a potential for radioactive releases in excess of 10CFR100, " Reactor Site Criteria", limits.
The allowable completion time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.
If the Reactor Coolant System pressure SAFETY LIMIT 2.1.5 is exceeded in MODE 3, 4, or 5, Reactor Coolant System pressure must be restored to within the SAFETY LIMIT value within 5 minutes. Exceeding the Reactor Coolant System pressure SAFETY LIMIT in MODE 3, 4, or 5 is more severe than exceeding this SAFETY LIMIT in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SAFETY LIMIT within 5 minutes. The action does not require reducing MODES, since this would require reducing ;
temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
If either SAFETY LIMIT in 2.1.A or 2.1.B is violated, the NRC Operations ;
Center must be notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in accordance with 10CFR50.72.
If either SAFETY LIMIT in 2.1.A or 2.1.B is violated, the Vice President Nuclear Generation shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period provides time for the plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to senior management.
If either SAFETY LIMIT in 2.1.A or 2.1.B is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC and the Vice President Nuclear Generation. This requirement is in accordance with 10CFR50.73.
If either SAFETY LIMIT in 2.1.A or 2.1.B is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. i l
l l
a B.3.1-2 3.1 REACTOR COOLANT SYSTEM Bases continued A. Operational Components (continued)
Reactor coolant pump start is restricted to RCS conditions where there is pressurizer level indication or low differential temperature across the SG tubes to reduce the probability of positive pressure surges causing overpressurization.
The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point.
These valves are considered OPERABLE at 3% of their setpoint of 2485 psig. Following testing the valve lift settings are restored within a nominal 1% of their setpoint. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F. The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load (Reference 1).
The requirement that two groups of pressurizer heaters be OPERABLE provides assurance that at least one group will be available during a loss of offsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions.
The pressurizer power operated relief valves (PORVs) operate to relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The PORVs are pneumatic valves operated by instru-ment air. They fail closed on loss of air or loss of power to their DC solenoid valves. The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses.
The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:
~~a. Manual control of PORVs to control reactor coolant pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown.~~
~~o B.3.1-3 3.1 REACTOR C001 ANT SYSTEM Bases continued A. Operational Components (continued)~~
~~b. Maintaining the integrity of the reactor coolant pressure boundary.~~
~~This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.~~
c. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item b. above).
~~d. Manual control of a block valve to isolate a stuck-open PORV.~~
The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS temperature is less than 310*F*.
The PORV control switches are three position switches, Open-Auto-Close. A PORV is placed in manual control by placing its control switch in the Closed position.
The minimum pressurization temperature (310'F*) is determined from Figure TS.3.1 1 and is the temperature equivalent to the RCS safety relief valve setpoint pressure. The RCS safety valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low temperature operational transients. Inadvertent pressurization of the RCS at temperatures below 310*F* could result in the limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded. Thus the low temperature overpressure protection system, which is designed to prevent pressurizing the RCS above the pressure limits specified in Figures TS.3.1-1 and TS.3.1-2, is enabled at 310*F*, Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limits of Figures TS.3.1-1 and TS.3.1-2 from being exceeded.
The setpoint for the low temperature overpressure protection system is derived by analysis which models the performance of the low temperaf.ure overpressure protection system assuming various mass input and heat input transients. The low temperature overpressure protection system setpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 and TS.3.1-2) are revised.
The 3 square inch RCS vent opening is based on the 2.956 square inch cross sectional flow area of a pressurizer PORV. Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.
~~Valid until 20 EFPY~~
m i
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B.3.4-1 3.4 STEAM AND POWER CONVERSION SYSTEMS Bases A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.
The ten steam generator safety valves have a total combined rated capability of 7,745,000 lbs/hr. The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten steam generator safety valves will be able to relieve the total steam flow if necessary (Reference 1).These valves are considered OPERABLE at i 3% of their specified setpoint. Following testing the valve lift settings are restored within a nominal 1% of their setpoint. ,
I In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be assured by availability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the steam generator safety valves. ,
One auxiliary feedwater pump can supply sufficient feedwater for removal of 1 decay heat from one reactv- The motor-driven auxiliary feedwater pump for I each reactor can be made available to the other reactor. During STARTUP l OPERATIONS, the Auxiliary Feedwater motor-operated injection valves maybe less I than full open as necessary to faciliate plant startup.
The minimum amount of water specified for the condensate storage tanks is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shutdown. Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling !
water system.
The two steam generator power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a steam generator tube rupture event (Reference 3) and following a high energy line rupture outside containment (Reference 2). The steam generator power operated relief valves are provided with manual upstream block valves to permit testing at power and to provide a means of isolation.
In order to assure timely response to a steam generator tube rupture event, a steam generator power operated relief valve is considered operable when it is capable of being remotely operated and when its associated block valve is open.
Isolation dampers are required in ventilation ducts that penetrate those rooms containing equipment needed for a high energy line rupture outside containment.
L_
4 b
B.3.4-2 l The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.
Re ferences
~~1. USAR, Section 11.9.4~~
~~2. USAR, Appendix I l~~
=
~~3. USAR, Section 14.5.4 f~~
a 4
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

ML20083B635

Text

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