Text

Tech Specs 1.23,2.1,2.2,3.1-3.10 & 4.1-4.10 Re Facility Operation to Be Modeled After B&W Sts.Encl to Util 750815 Ltr
	ML19329A993
Person / Time
Site:	Davis Besse
Issue date:	08/15/1975
From:	; TOLEDO EDISON CO.
To:	;
References
	NUDOCS 8001280742
	Download: ML19329A993 (133)
	v • d • e

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~ 1.23 SHIELD BUILDING IlfrEGRITY shall exist when ,J;-

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1.23.1 The airtight doors and the blowout panels listed in Table 4.6-xx are closed except when the airtight doors are being used for normal transit entry and exit.

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1.23.2 The Emergency Ventilation System is OPERABLE.

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' - OPERATIONAL MODES

'i$ REACTIVITY % RATED AVERAGE COOLANT

@p' _

CONDITION, Keff TEMPERATURE l l@DE, THERMAL POWER *

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1. POWER OPERATION _

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% 3. HOT STANDBY 0 - 280*F>Tavg > 200"{

4. HOT SHUTDOWN <0.99

<0.99 0 1200*F

5. COLD SHUTDOWN 1095 0 1140*F [

6. REFUELING **

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~ Excluding decay heat. nbolted or removed and fuel in the vessel.

- Reactor vessel head C. . .M,,

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1-6 June 24,1975 i DAVIS-BESSE, UNIT 1 ,

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52fd_%5. .f...$$-..;,..

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- .- PROOF & REVIEW CO. y U 2.0 SAFETY LIMITS AND LIMITIfiG SAFETY SYSTEM SETTIflGS

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~7 2.1 SAFETY LIMITS

M1

., REACTOR CORE 2.1.1 The combination of the reactor coolant core outlet pressure and

. ,(*gh outlet temperature shall not exceed the safety limit shown in Figure gg

-m 2.1-1.

M.d APPLICABILITY: MODES 1 and 2.

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C "9 Whenever the point defined by the combination of reactor coolant core

, outlet pressure and outlet temperature has exceeded the safety limit.

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E M If$ % } A Y M S$$ N I@ 5 M 85306d M k l PROOF & REVIEW COPY V SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

'J*Ia REACTOR CORF ;

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1 2.1.2 The combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE

. ws shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of two, three and four reactor coolant pump operation.

'g APPLICABILITY: MODES 1.

~

g+M y{L. ACTION: -

~

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N. Whenever the point defined by the combination of Reactor Coolant System 2 flow, AXIAL POWER IMBALANCE and THERMAL POWER has exceeded the appropriate

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AXIAL POWER IMBA' AllCE % . .

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. . CURVE REACTOR COOLANT FLOW (Ib/hr) .

0

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. 2 98.1 x 108 :

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. i t . e REACTOR CORE SAFETY LIMIT 1.'

. Figure 2.1-2 .

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REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPCINTS l). 3 $.

TRIP SETPOINT ALLOWABLEYklUES FUNCTIONAL UNIT -

Not Applicable Not Applicable *

(

1. M,anual Reactor Trip . .

% y -

$105A4 % of RATED THERMAL POWER

. $ los.so'f RATED THERMAL POWER l

2. Nuclear Overpofer g '

RCS Outlet Temperature-High < 618.9f F '

<619F

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$ e 3.

?g( m: quis JMsM Axst As aw, , n

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4. Nuclear Overpower Based on g'f

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g ; ._

% ** 2 /- / $ $ *_'i.N Bi ? - 4 0. 0 RCS Flow and AXIAL POWER -

S2.I - 2_A.o

. IMBALANCE (2) g,, i _39 5zg,75 * . %6 + %.O g3 4 + g S,75

.g 0+ 6 +" 19 d 5s to m

& p{c

E. /,06

? ..

" W When the Nuclear Overpower trip shNint"Is requir(d to be reduced by an ACTION statement. ' '

. 'to some percentage'of the THERMAL POWER" allowable for the rea'ctor coolant, pump combination.

the THERi*.AL POWER allowable: .

~* For 3 pump operation,.'is 80.7% of RATED THERMAL POWER.

i a. -

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b. For cperation with one pump in each loop is $3 % of RATED THERMAL POWER.

~ ~ ~ - ' - ~ ~ -

l

~

(2) Trip nay be manually bypassed by actuating Shutdown Bypass provided '

- that by administrative control the Nuclear Overpower trip setpoint '

A Shutdown ,

i- 3 is reduced to a value < 5% of RATED 'mERMAL IWER._

3

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. Bypass RCS pressure-High trip setpoint of.1820 psig is automatically

." imposed when the Shutdown Bypass is actuated thus actuation of the .

i 'S Shutdown Bypass above 1820 psig will produce a reactor trip. .

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, . . TABLE 2.2-1 (Contraurd) .v

% ~ '

. REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS -

((h Q .

1 h FUNCTIONAL UNIT ,

TRIP SETPOINT ALLOWABLE VALUES 1;

5. RCS Pressure-Low (2) 11985.4'osig

1 196 5'psig ,

6. RCS Pressure-High 1 Z,=.,55 psig

{- , 5 7.354.6 psigf .

. 7. RCS Pressure-Variable Low 2) 1((13 85)Tout ' .G4 Psig ' , (13.%) Tout ~ N 9

.Psig

8. huclearOverpower. Based 1 /25,0for RATED WE ." 1 N8 of RATED DN on Pump Monitors (2)'

~

POWER with three pumps operating POWER with three pumps operating 7

(2/1 or 1/2) . (2/1 or 3/2)

-l ,. -

1.f4,38%of RA'IED TIERMAL 1 f.f O %,of RATED UIERMAL -

. POWER with one pump operating POWER with one punp operating

. in each loop (1/1) -

in,each loop (1/1)

~

1 0.0% of RATED THERMAL *1 Ng of. RATED MIERMAL ,

POWER with two pumps operating' POWER with two pumps operating.

3" . in cne loop and no pumps in one loop and no pumps

. ~1, . .

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operating in the other loop operating in the other loop U -

(2/0 or 0/2) (2/0 or 0/2) h , d o o% of RATED DIEBMAL

1 NS of RA7ED TURIAL POWER with no punpa operating P0kJER with no punps operating 4 '

k or only one pump operating. or only one punp operating.

(0/1,.1/0, or 0/0) . (0/1, 1/0, or 0/0) .

9.*

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. 540 56 0 580 600 620 640 lieactor Outlet Temperature, F

- ~-

.e.

c.- .

+

, PROTECTI VE SYSTEM A AXIHUM ALLOWABLE SET POINTS.

, .g

~.

PR ESSUR E/ T EMP ER A_T_U._R_E. _L_i h.i T_S .A..T._MA_X_.I MUM ALLO 11ABLE l'0!;LR FOR MittlMUM D.'GR e .- .

. FIGUR.E 7. 2 -2 (

MDMt;:stS- 2-10 . .

M- / .

Y -

!, c 2.1 SAFETY LIMITS PROOF & REVEW COPY i BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overhetting of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation l temperature. l Operation above the upper boundary of the nucleate boiling regime vould result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-sture and Pressure have been related to DNB through the B&W-22W-33 DNB e- :orrelation. The DNB correlation has 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, DNBR, defined as the ratio of the neat 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 DNBR during steady state operation, normal

)perational transients, and anticipated transients is limited to 11<30). /.3 a This value corresponds to a 1957 percent probability at a 199h percent

onfidence level that DNB will not occur and is chosen as an appropriate nargin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at 1 mirimum DNBR of (1.32/J440-) is predicted for the maximum possible thermal

)ower 112% when the reactor coolant flow is 1131.3 x 106X 1bs/hr, which is the design flow rate for four operating reactor coolant pumps. This l :urve is based on the following nuclear power peaking factors with potential

Fuel densification effects:

l . .

l , '2 56

j 71 FN = (E<67); N F3H , (L*73.); pN=fl.50Y rhe design limit power peaking factor are the most restrictive

alculated at full power for the range from all control rods fc'?y-withdrawn to minimum allowable control rod withdrawal, and form ti..

4

ore DNBR design basis.

i

. DAVIS-BESSE, UNIT 1 B 2-1 June 24, 1975 .

4 N

I I

9

1 PROdF & RiMEW COPY'

(' SAFETY LIMITS BASES

- The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:

1. The'(1.32/M DyBglimit produced by a nuclear power peaking factor of FN = (2 4 7J or the combination of the radial peak, L

axial peak and position of the axial peak that yields no less than a (1.32/1 30r) DNBR.

2. The combinatinn of radial and axial peak that causes central .

fuel melting at the hot spot. The limit is (1987 kw/ft.

20.4-Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2

orrespond to the expected minimum flow rates with four pumps, three pumps, l and one pump in each loop, respectively. .

- The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1. The curves of BASES Figure 2.1 represent.the conditions at ahich a minimum DNBR of (1.32/1307 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to (22#5)%,

j shichever condition is more restrictive.

Using a local quality limit of (22/J&T% at the point of minimum

)NBR as a basis for curve 3 of BASES Figure 2.1 is a conservative criterion aven thougn the quality c.t the exit is higher than the quality at the Joint of minvaum DNBR.

The DNBR as calculated by the (B&W-2A<I) DNB correlation continually increases from point of minimum DtjBR, so that the exit DNBR is always ligher. Extrapolation of the correlation beyond its published quality range of (+22C5)S is justified on the basis of experimental data.

The maximum THERMAL POWER level during partial pump operation is flow dependent and limited by a power level trip produced by the flux / flow ratio plus the maximum calibration and instrument errors. For ext ple, it 3 pump operation.with E sflow, the maximum THERMAL POWER is (64:5)%

(JS)% flow x (.14415) = (.78)% indicated power. 7z2v 74.7 7, /d WJ & ' 7+.+ ~4

) AVIS-BESSE, UNIT 1 B 2-2 April 23,, 197S 4

pmi l i

l WSflMkMMMNEM2Ed$NEESMNND , .

\

4

. r PROOF & REVIEW CO LIMITIflG SAFETY SYTEM SETTIflGS

'W R BASES q .,f I t'4 f,

l(uclear Overpower Based on RCS Flow and AXIAL POWER IMBALAtlCE

.. .M C

N'i

-d:M The power level trip setpoint produced by the reactor coolant system flow is based on'a flux-to-flow ratio which has been established to accomodate flow decreasing transients from high power where protection

($1

a is not provided by the fluclear Overpower Based on Pump Monitors channels. -

7,,,$,d o-The power level trip setpoint produced by the flux-to-flow ratio "i 3rovides both high power level and low flow protection in the event the

- ceactor power level increases or the reactor coolant flow rate decreases.

The power level setpoint produced by the flux-to-flow ratio provides a )verpower Df18 protection for all modes of pump operation. For every flow ate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as i, follows: ,

V l- P

}i would occur when four reactor coolant pumps are operating power if 108.0% and reactor flow rate is 100%, or flow rate is 92.6% and power level is 100%.

. {t

% 2.

Trip would occur when three reactor coolant pumps are operating if power if 80.7% and reactor flow rate is 74.7%, or flow ~

rate is 69.4% and power is 75%.

1 3.

Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.9%

% and reactor power level isflow rate is 49.0% or flow rate is 45.4% and the 49.0%.

e rrors For safety calculations the maximum calibration and instrumentation for the power level were used.

The AXIAL POWER IMBALANCE boundaries are established in order to a

prevent reactor thermal limits from being exceeded. These thermal limits re either. power peaking kw/ft limits or DNBR limits. The AXIAL POWER s

MBALANCE reduces the power level trip producqd by the flux-to-flow ratio

' uch that the boundaries of EASir Figure 2.2'are produced. The flux-to-r low ratio reduces the power level ' trip and associated reactor power-eactor power-imbalance boundaries by 1.08% for a l% flow reduction.

v

,. 1 UAVIS-BESSE , U IT 1

' B2-5 April 23, 1975

{

o

\

i LIMITING SAFETY SYSTEM SETTINGS PROOF & REVIEW COPY BASES Swn in P;5a re 2.2 -2 i /

RCS Pressure - Low, High and Variable Low l During a slow reac ivity insertion startup accident from low power or a slow reactivity insertion from high power, the RCS Pressure-High setpoint is reached before the Nuclear Overpower trip setpoint. The trip setting limit /for RCS Pressure-High 2355 psig has been established to maintain the system pressure below the safety limit 2750 psig for any design transient. The RCS Pressure-High trip also backs up the Nuclear Overpower trip. 54 e , 2.2-2 .

IWg A cS ,73 M11 The(kCS Pressure-Low 1900'psig an f

RCS P

,ressuren Variable Lowc.

Tout *F- 287f psig trip setting limits have been established to maintain the DNB ratio greater than or equal to (1.32/,L30")'for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.

Dere- tu i.-he-calibration--and-instrumentation-errors,-the-safety-t . anal-ysismsed-a- RCS Pressure =Vartable-tow-t:-i;r setpcint of-115 25-Tout T- (7923-psig). .

Nuclear Overpower Based on Pumo Monitors In conjunction with the power / imbalance / flow trips the Nuclear Overpower Based On' Pump Monitors trip prevents the minimum core DNBR fron decreasing below (1.32/1.30) by tripping the reactor due to the loss of reactor coolant pump (s). The-pump monitors One#

, 3 also restrict the power level for a s a O' the number of pumps in operation. TAis trip Reactor Containment Vessel Pressure - High .

The Reactor Containment Vesstl Pressure-High trip setting limit

< 4 psig provides positive assurance that a reactor trip will occur in l the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure -Low trip. ,

Ghat J~ 6ypass Sa.&wu,Q .

DAVIS-BESSE, UNIT 1 . B 2-6 June 24, 1975 l

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,].D Shutdown Bypass g ...i n)

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- In order to provide forecontrol rod drive tests, zero power physics testing, and-startup proceduresj thereis provision for bypassing certain segments of )! I the reactor protection system. The reactor protection system segments which .* !

-can be bypassed are shown in Table if.2-/ Two conditicus are imposed when . -

. . .c -

.* .&.'

C '

the bypass is used: )_._ l' M. .S '

.I

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1. By administrative control the nuclear overpower trip set point must be .;g.:

reduced to a value <.5.0% of rated power during reactor shutdown.

. - 5: t;:je

r. .

2. A,high reactor coolant system pressure trip setpoint of /6J'O psig is .

i ;gq automatically impos,ed. . i.3.; '

- .:.. . O,A. .g.

h purpoce of the/S2o psig high pressure trip set point is to prevent normal cperation with pr.rt of the reactor protection system bypassed. This high -l Pressure trip set point is lower than the normal low pressure trip set point -

pj so that the reactor must be tripped before the bypass is initiated. The 7 1 over power trip set point of 15.0% prevents any significant reactor power '

from being produced when perforning the physics tests'. Sufficient natural '

circulation (5) would be available to remove 5.0% of rated power if none of i ths reactor coolant pumps were operating. -

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' 'i 6 20 640 660 580 600 . gi

+ %

Reactor Outlet Temperature, F .

.i .

1

. . REACTOR COOLANT .

. FLOW POWER PLMPS OPERATlHG (TYPE OF LlHIT)

-- . ..2 . . .. CURYE (LBS/HR)

'i 131.3 x 106 ( 100 %) I12% FOUR PUMPS (DHBR LlHIT)

..'N' 2 98.1 x 106 (74.7%) B77 THREEPUMPS(DNBRLlHIT)

~ .

597 ONE PUMP IN EACH LOOP M* .

3 611.14 x 106(49.0%)

, (QUALITYLlHIT) -l

. 1 s, .

f i 4 4 CORE PROTECTION SAFETY LlHITS I

~

j gg Figure Li l .

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82-6 .

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@ M M5EMIFa3IE % fM Mf3MMfD M5NEN M 8N

- j l

( PROOF & REVIEW Con

." h 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 4% BORATION CONTROL

-T- r SHUTDOWN MARGIN

'. r,

m. N 5 Y

~J ;#' LIMITING CONDITI0il FOR OPERATION

). .

Qz;

% 3.1.1.1 The SHUTDOWN MARGIN shall be > 1% ak/k.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

Qg

- Q* ;f

~~L *

.M. TION:

/

/W \

M 1 Hith the SHUTQOW[N boration a MARGIN gpm of ($pm < 1% ak/k,equivalenruntil boro'n,~~or'it's immediately initiate the andl conl required SH OWiM4ARii1H 1s restored.

I

! SURVEILLANCE REQUIREMENTS (

4.1.1'.1.1 The SHUTDOWN MARGIN shall be determinsd to be > 1% ak/k:

b a. When in MODES 1 or 2, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , by verifying that control groups withdrawal is within the limits of Specifi-l I cation (3.1.3.6).

.: d When in MODES 3, 4 or 5, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by b.

consideration of the following factors:

1. Reactor coolant system boron concentration, i

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on grcss thermal energy generation,

. 5. Xenon concentration, and l

7- .

6. Samarium concentration. '

I

c. Immediately upoi. detection of hn immovable or untrippable inoperable control rod, considering the worth of the immovable

- or untrippable inoperable control rod as well as assuming that the single OPERABLE rod of highest reactivity worth is fully

, withdrawn.

.. M Immediately upon detection of a moveable and assumed trippable d.

inoperable control rod, assuming that the single rod of highest reactivity worth is fully withdrawn. l s e. Prior to initial operation above 5% RATED THERMAL POWER after I

' each refueling, with the control groups at the minimum with- l I

drawal limit of Specification (3.1.3.6). l CiVIS-BESSE, UNIT 1 3/4-1-1 June 24, 1975

~ '

L.

^

REASON 18 3.1.1.2 Delete the parenthesis from 2800 gpm. The bases for this spec assumes a turnover of the RCS inventory in

( 30 minutes. The following calculation is the bases for 2800 G.P.M.:

RCS volume . L = Desired DHR flow

, :" 30 minutes (11,262 ft.3) (7.48 eal)_ 2 1 gg J : 30 minut.ss : Desired DHR flow

.84239.76 .*. 1 30 minutes = Desired DHR flow 2807.992 gpm = Desired DHR flow 2800 GPM 7 esired D DHR flow .

NOTE: 11,262 ft3 is the total water inventory in the RCS as shown on DB-1 FSAR Figure 5.2.

4,g, g, p Delete "with a discharge flowrate of > (5000) gpm and" The operator will be reanug the flowrate "to the core" on the instrumentation in the control room. Below is a simplified drawing of a DHR pump.

,l,REsTmcried estance

. 'y ,

' l' - Tb THE Q

'I Res

. gy com gruom.u n,.n usa arica osc0GPn 1

R 9 8 i

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.. 1 1

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p.. -

V REACTIVITY CONTROL SYSTEMS PROOF & liEVIEW COPY 1';*

BORON DILUTION

.m .. .

LIMITING CONDITION FOR OPERATION p ,.y.'

@$ 3.1.1.2 The Reactor Coolant System flow rate to the core shall be l

> 12800) gpm whenever a reduction is being made in the boron concentra-i.$,:

LW tion of the Reactor Coolant System.

$Pv. '

7, y APPLICABILITY: ALL MODES. ,

M ACTION:

~

With the Reactoi- Coolant System flow rate to the core < 42800h gpm, l

t. ' J imediately suspend all operations involving a reduction in boron

fi.e O concentration of the Reactor Coolant System.

l y. '

l l IQ -

gj SURVEILLANCE REQUIREMENTS

'fJ%d

.o 4.1.1.2 The reactor coolant flow rate'to the core shall be j determined to be > .('28001 gpm prior to the start of and at least once J per hour during a reduction in the Reactor. Coolant System boron concen-tration by either:

W a. Verifying at least one reactor coolant pump is in operation, or

^ ~

l b. Verifying that at least one DHR pump is in operation (gdth a

! dischargeJ10wrate-of->-(5000}--gpmjand is aligned to supply l > $2800)".gpm flow to the core.

_ l M. - .,' .

v

~l DAVIS-BESSE, UNIT.1 3/4 1-3 June 24, 1975

W & [?N Q h S Y K 4 % W ? $ $ @ N W $ % 5 T l -

. r-

%s5 D .

REACTIVITY C0flTROL SYSTEMS o

PROOr" a kckW "

COP.y M

MINIMUM TEMPERATURE FOR CRITICAlli.

y,* , LIMITING CONDITION FOR OPERATION ,

s: 2 /. /. 4-

.i A-bl4- The RCS temperature (Tavg) shall be > 525 *F whenever the 8.N.

e-reactor is critical 1

f- .

. APPLICABILITY: MODES.1 and 2 and #

m .

ACTION: -

7, e. . f m

Wi WithRCStemperature(Tavg) < 525. *F, trip the reactor.

7 SURVEILLANCE REQUIREMENTS -

4 .1.1.4 The RCS temperature (Tavg) shall be determined to be > 525 F:

,i Q ,

a. Within 15 minutes prior to making the react.or critical; Yd Ew3 b. At least once every hour when RCS temperature (Tavg) is

< 532 *F.

y With Keff > 1.0.

S ee Special Test Exception 3.10.2.

l I

w -

. 3 .

I v -

DAVIS-BESSE,, UNIT 1 3/4 1-6 ,

June 24, 1975

.e

y g t,; 5,/ yij,/ { Q [.'-[;

i

! REACTIVITY C0tlTROL SYSTEMS 3/4.1.2 80 RATION SYSTEMS

FLOW PATHS - S!!DTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE.

a. A flow path from the boric acid addition system via a boric '

acid pump and a makeup or decay heat removal (DHR) pump to the Reactor Coolant System, if only the boric acid storage system in Specification 3.1.2.$5 is OPERABLE, or

b. A flow path from the borated water storage tank via a makeup or DHR pump to the Reactor Coolant System if only the borated water storage tank in Specification 3.1.243b is OPERABLE. ,

l APPLICABILITY: MODES 5 and 6. ,

. ACTION:

With none of the above. flow paths OPERABLE, suspend all operations i involving CORE ALTERATIONS or positive reactivity chan'ges until at least one injection path is restored to OPERABLE status, f

SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demon-

. strated OPERABLE:

i

a. At least once per 7 days by:

1. Exercising all testable power operated valves in the flow path required for boron injection through at least one

. complete cycle, *

2. Verifying that the pipe temperature of the heat traced portion of the flow path is above 105 F if only a flow path from the boric acid addition system is OPERABLE, and i

b. At least once per 31 days by verifying the correct positinn of O all manually operated valves in the boron injection flow path not locked, sealed or otherwise secured in position.

DAVIS-GESSE, UNIT 1 3/4 1-7 April 23, 1975

I . ..

j

'4 PROOF & Ri:VlEW COPY REACTIVITY CONTROL SYSTEMS

- BORATED WATER SOURCES - SHUTDOWN-

.n.

QQ.S f LIMITING CONDITION FOR OPERATION

?.4g .

$$4 is a minimum, one of the following borated water sources shall y

Ifd 3.1. 2. 8 be OPERABLE:

a. A boric acid ad ~ tion system and one associated d.' heat tracing ci cuit with:

.Nt

.G9sogallons, M 1. A minimum contained volume of

--q

! z o

2. Between$7#7and /2j ppm of boron, and l' ' 3. A minimum solution temperature of (105) F.

l -

b. A borated water storage tank (BWS1) with:

l

1. A minimum contained volume of 36CT,000 gallons, equ'ivalent

.i to a level of 28 feet, M 'Q

2. A minimum baron concentration of 1800 ppm, and 43 M --- 3. A minimum solution temperature of 35 F.

APPLICABILITY: MODE 5.

4 ACTION _:

With no borated water sources OPERABLE, suspend all op'erations involving q positive reactivity changes until at least one borated water source is restored to OPERABLE status.

+Eq l

SURVEILLANCE REQUIREMENTS l .

4.1.2.8 The above required borated water source shall be demonstrated' l

l OPERABLE:

a. At least once per 7 days by:

l .,

1. Verifying the boron concentration of the water, v 2. Verifying the water level of the tank, and

y. April 23, 1975 3/4 1-15' DAVIS-BESSE, UNIT 1 3

.a Mdf8N8$$Nd5N$$DME5EfE$k!8EMN!GM5

- *6 PROOF & REVIEW COPY REACTIVITY CONTROL SYSTEMS

, e

~w- BORATED WATER SOURCES - OPERATING k$

LIMITING C0f!DITION FOR OPERATION k..$Y . .,

.-[s 5 3.1.2.9 Each of the following borated water sources shall be OPERABLE:

>..+ SU*i ;;

A boric acid addition sy' stem and one associated heat tracing Q.) -

a.

circuit with:

ar.w gallons,

- 1. A minimum contained volume of

2. . . Between F7e7and /4./nppm of boron, and hI 3. A minimum solution temperature of. (105)*F.

I b. A borated' water storage tank (BWST) with:

1. A minimum contained volume of 360,000 gallons of water,

- equivalent to a level of 28 feet, ID 2. A minimum boron concentration of 1800. ppm, and g,

'3. A minimum solution temperature of 35'F.

APPLICABILITY _: MODES 1, 2, 3 and 4.

ACTION:

~

a. With the boric acid addition system tanks inoperable, restore the storage system to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months e .

I or make the reactor subc-itical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a SHUTDOWN MARGIN' equivalent to '1% ak/k ~at 200 F;

' restore the concentrated boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . ,

b. With the borated water, storage tank inoperable, restore the tank to OPERABLE status within one hour or be in COLD SHUTDOW l

l

.e a within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t DAVIS-BESSE; UNIT 1 . 3/4 1-17 April 23,1975 t

$NINMEN @ N M kT@'$FM56' EMU 6DisNM5h 1

' PROOF & REVIEW COFY V

REACTIVITY CONTROL SYSTEMS

11. 4 7- ie SURVEILLANCE REQUIREMENTS

. eft i- 4.1.2.9 Each borated water source shall be demonstrated OPERABLE:

-afei a. At least once per 7 days by:

q p$hN,% Verifying the boron concentration in each water source, wd 1.

rug

2. Verifying the water level in each water source, and .

.h[f Verifying the boric acid addition system solution

3.

~

1, temperature.

' b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the BWST temperatu~re

'd r . when the BWST ambient air temperature is < 35 F.

l i

V M

M = 4 i

t.

M h e r i

v .

~

May 19, 1975 DAVIS-BESSE, UNIT 1 . 3/4 1-18 t *

j

N -

A *

- ~

7..

REACTIVITY CONTROL SYSTEMS PROOF & REVIEW COPY 3/4.1.3 MOVABLE C0flTROL ASSEMBLIES .

T GROUP HEIGHT - SAFETY AND REGULATING ROD GROUPS

%k

. we,7 LIMITING CONDIT"N FOR OPERATIONS h

o

u. ' 3.1. 3.1 All control (safety and regulating) rods shall be OPERABLE and u, positioned within 16.5% (indicated position) of their group average g#)c

.w height. .

.. a 13Mh APPLICABILITY: MODES 1* and 2*.

MM^

ACTION:

.- With one or more control rods inoperable due to being immovable a.

9

~

as a result of excessive friction or mechanical interference or' known to be untrippable: ,

B

1. Immediately determine that Specification 3.1.1.1 is satisfied, and D 2. Be in HOT STANDBY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

h&' b. .With a maximum of one control rod inoperable (unless immovable as a result of excessive friction or mechanical interference or

-~ knoun to be untrippable) or misaligned from its group average height by more than i 6.5% (indicated position), operation may continue provided that:

1. With an inoperable control rod;

~

y a) . Specification 3.1.1.1 is satisfied, and All other control rods not fully inserted are demon-b b) strated OPERABLE by movement of at least L5%'at

. least once every 14 days. 2%

2. With a misaligned , control rod, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; a) The misaligned rod is returned to within the above j

alignment requirement, or l

]-

i. b) The THERMAL POWER is reduced to < 60% of the

~

.TilERMAL POWER allowable for the Teactor coolant pump combination, except that, a

See Special iest Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 l~19 May 27, 1975 s

@ M M .$ $ N k M 5$ $ $ $ N.. S$$ $ G S 3 ki M N I4 N Nk

.l

. .~ j

-; D -

PROOF & Ri:VlEW COPY REACT'.7ITY CONTROL SYSTEMS

.r a ,

T ACTION: (Continued)

, f, n

, g c) For a misali.gned regulating rod,-operation above j (60)% of the THEPJ4AL POWER allowable for the reactor r% . coolant pump combination may continue provided that, R"f.fi

,4e The group is positioned such that the misaligned 1)

G_*/fi regulating rod is restored to within the limits W for the group average height, and

'idd

$jt 2) The withdrawal limits of Specification 3.1.3.6 g ,

are satisfied.

3. An analysis of the potential ejected rod worth is per-

Wy formed within 3 days and the rod worth is determined to be

. < T0.657% ak/.k at RATED THERMAL POWER and < $1.0}% Ak/k at

^

hot zero power during the remainder of the fuel cycle.

I

.., SURVEILLANCE REQUIREMENTS f

%"q~

4.1.3.1.1 The position of each control rod shall be determined to be within the group averace height limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals wnen the Asymmetric Rod Fault Circuitry is inoperable, then verify the rods with the inoperable fault circuitry to be within the limt at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3~'J 4.1. 3.1. 2 Each control rod not fully, inserted shall be determined to be

- T OPERABLE by movement of at least .2.E at least once every 31 days.

h 2. X i N'.

l l

DAVIS-BESSE, UNIT 1 3/4 1.20 May 2, 1975 l

h

. _ . . . . .g

15i4WMvM#NPM@%RXM:WMW2ERMf$$ -

~ .

PROOF & REVIEW COia'I REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - AXIAL POWER SHAPIllG R00 GROUP g

LIMITING CONDITION FOR OPERATION

'd41 Y '

3.1.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless Jed7,1

'W3 fully withdrawn, and shall be position.ed within + 6.5% (indicated position)

~

jg of their. group average height.

APPLICABILITY: MODES 1* and 2*. .

d ACTION:

With a maximum of one'APSR inoperable or misaligned from its group c" average height by more than + 6.5% (indicated position),. operation may continue provided that within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

w I a. The APSR group is positioned such that the misaligned rod is restored to within limits for the group average height, or j V b. It is determined that the imbalance limits of Specification 3.2.1 are satisfied and movement of the APSR group is pre-vented while the rod remains inoperable or misaligned.

u .~ .. .

- SURVEILLANCE REQUIREMENTS 4.1. 3 . 2.1- The position of each APSR rod shall be determined to be within the group average height limit by verifying the individual rod ed positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when ic& the Asymmetric Rod Fault Circuitry is inoperable, then verify the rods with the inoperable fault circuitry to be within the limit at least

$M' once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l 4.1.3.2.2 Unless all AP3R are fully withdrawn, each APSR shall be determined to be OPERAGLE by moving the individual rod at least JefDf l '

2.o z l at least once every 31 days.

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 1-21 June 4, 1975 t

v 4

e

@ ?sM2ENINf7IY$ @ dif M @ $6IM:D$Dd'5@5$f M $

! l

d. PROCF & REVIEW CO.-Y REACTIVITY C0flTROL SYSTEMS POSITI0ft INDICATOR CHANNELS iy

..:c,s

. v.,$M

LIMITING CONDITION FOR OPERATION

? .1 N l* " " a b s olv +c ki: .

f All safety, regulating and axial power shaping control rod--reed 3.1.3.3 f gg:1 switch position indicator chanriels and pulse %pping position indicator

'.i- j] channels shall be OPERABLE and canable of determining the control rod positions within + k-)f.

gyp *

& s 'P)

<v WN APPLICABILITY: MODES 1 and'2.

G;32 g /, j , W

'l ACTION _:

~ a'. With a maximum of one reed-switch-pasTtion indicator channel per control rod group or one pulsedtepping position indi::ator channel per control rod group inoperable and the control rod (s)-

with the inoperable position indicator channel partially inserted, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either: ,

1. Restore the inoperable position indicator channel to

,D OPERABLE status, or

. .g 2. Be in H0T STANDBY, or

3. Reduce THERMAL POWER to < MON of the THERMAL POWER allowable for the Reactor Coolant Pcmp combination.

- Operation at or below this reduced THERMAL POWER level may continue provided that within the ncxt 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a) The control rod group (s) with the inoperable position indicator is inserted or. withdrawn while maintaining ct^ ~ the group withdrawal sequence required by Specification (3.1.3.6), and when this group reaches

-g its fully or its 25%, 50%, or 75% withdrawn position, the " Full Out" limit or the position reference indicator channel of the control rod with the inoperable position indicator is actuated and verifies the position of this control rod. The THERMAL POWER level may be then returned to a level consistent with'all other applicable specifications; or g The control rod group (s) with the inoperable position M b) indicator is fully inserted, and subsequently main-

.tained fully inserted. while maintaining the withdrawal sequence and TilERMAL POWER level required by Specification (3.1.3.6) and when this control rod group reaches its fully inserted position, the " Full In" t

l DAVIS-BESSE, UNIT 1 3/4 1-22 gune4,1975 i -

s

s

y s

I REACTIVITY CONTROL SYSTEMS _ PROOF & liiNIEW COcY POSITION INDICATOR CHANNELS (Continued) i LIMITING CONDITION FOR OPERATION (Continued) limit of the control rod with the inoperable position indicator is actuated and verifies this control rod to be fully inserted. Subsequent opera;. ion shall be within the limits of Specification (3.1.3.6).

g .,g s e 7. rc ta rs e

b. With a maximum of one reed-switch position indicator channel per group or one pttise-stepping position indicator channel per group inoperable and the control rod (s) with the inoperable -

position indicator channel at either its fully inserted po-sition or fully withdrawn position, operation may continue provided:

1. The position of this control rod is verified immediately and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its " Full In" or " Full Out" limit (as applicable),

l ,

2. The fully inserted control rod group (s) containing the inoperable position indicator channel is subsequently maintained fully inserted, and .

Subsequent operation is 'within the limits of Specification 3.

%3.1.3.6).

rela tive

c. With one or more pulse 4tepping position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided all of the reed = switch position indicator channels are OPERABLE. gg SURVEILLANCE REQUIREMENTS 4.1.3.3 . re /a t n g g ,,

a. Each r:eed -swi-tch and-pulsentepping position indicator channel shall be determined to be OPERABLE by verifying that the ptrise tekf.

stepping position indicator channels and the reed switch position indicator channels agree within (2.)% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Asymmetric i Rod Monitor is inoperable, then compare the pulse stepping /c-lah W r

position indicator and reed-switch. position indicator channels at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . n 6se ta r <

l DAVIS-BESSE, UNIT 1 3/4 1-23' June 4, 1975

' )

n !.

3

PROOF & REVIEW CO. Y p REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued) i i LIMITING CONDITION FOR OPERATIONS (Continued)

b. Each position reference indicator channel shall be determined to be OPERABLE by verifying that; bsol 4*

1. Each position're$erence. indicator cht.nnel agrees with (2)%

of the reed switch position indicator channel at (25, 50 &

75)% withdrawal during each startup unless performed in the previous 7 days. .

2. The positio'n reference indicator channel of the control rod with the inoperable position indicating channel agrees within ( )% of the OPERABLE reed (--switch or pu-1se-positionindicatorchannelat1astonceper4hou{s.

a bsolu k relat ve t .

i DAVIS-BESSE, UNIT 1 3/4 1-24. June 4, 1975 l

. 3 l I(

th.f bC /

1 f .

( ~.

REACTIVITY CONTROL SYSTEMS _ PROOF & Ri-VIEW COPY

-A ROD DROP TIME N,

.- m,

.y'jSK LIMITING C0flDITI0fl FOR OPERATION M3

\h,h

? 3.1.3.4 The individual (safety and regulating rod drop time from fully 3.f.Ni/ withdrawn shall be < .1466' seconds from power interruption at the con-P's trol rod drive breaFers to 3/4 insertion (25% position) with:

a. > f500)*F, and T

avg .

r a

All reactor coolant pumps operating.

g ___.

b.

APPLICABILITY: MODE 3.

, ~. ;- -

ACTION:

j .m

' a. With the drop time of any safety or regulating rod determined l to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding t.o MODE 1 and 2.

b. With the rod drop times within limits but measured at less that

g -

full reactor coolant flow, operation may proceed provided that THERMAL POWER is restricted to less than or equal to the THERMAL

- POWER allowable for the reactor coolant pump combination opera-

.r ting at the time of ' rod drop time measurement.

SURVEILLANCE REQUIRE'MENTS 7 4.1.3.4 The rod drop time of safety and regulating rods shall be demon-

['eL, strated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any main-tenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

.+

l c. At least once every 18 months.

June 4, 1975 DAVIS-BESSE, UNIT 1 3/4 1-25 4

6 6

i d

/~' REACTIVITY CONTROL SYSTEMS R0D PROGRAM i

\

j LIMITING CONDITION FOR OPERATION t

3.1.3.7 Each~ control rod (safety, regulating and APSR) shall be programned l to operate in the core position and rod group specified in Figures (3.145) u ~'

and (3.1-5). 3. / APPLIC/BILITY: MODES 1*, 2* and 3*.

I ACTION:

With any control rod not , rogrammed to operate as specified, program the rod to operate as specified prior to proceeding to MODES 1, 2 or 3.

SURVEILLANCE REOUIREMENTS 4.1.3.7

a. Each control rod shall be demonstrat'ed to be programmed to operate in the specified core position and rod group by:

1. Selection and actuation from the control room and verifi-cation of movement of the proper rod as indicated by both the absolute and relative position indicators:

a) For all contr)1 rods, after the control rod drive patchs are locked subsequent to test, reprogramming or maintenance within the panels, b) For specifically affected individual rods, following maintenance, test, reconnection or modification of power on instrumentation cables from the control rod drive control system to the control rod drive.

2. Verifying that each cable that has been disconnected has been properly matched and reconnected to the specified control rod drive.

b. Control rod drive patch panels shall be locked with limited access to be authorized by the Station Superintendent.

Sce Special Test Exception 3.10.1.

l DAVIS-BESSE, UNIT 1 3/4 1-32 July 3, 1975 i

e.

J G 7 6 .

3 2 1 4 5 8 5 .8 5 ,

. . 4 7 , 7 3 i

l 6 .8 5 6 5 8l 6 -

.1 ,2 2 2 .

l> S 6 7 6 5 7 2 2, 2, .1 6 8 5 6 5 8 6 3 7 7 4 5 8- 1 S 8 - 5 ,

4 1 l2 3 .

. 6 7 6 -

v . FORILLUSTRATI0ftONLY .

BANK H0. RODS PRUPOSE

1 4 Safety -

2 8 Safety -

3 4 Safety

4. 4 Safety 5 12 , Regulating 6 12 Regulating .

7 9 Regulating .

8 8 APSR CONTROL R00 C07,E LOCATION AND GROUP ASSIGH:$1TS_:5 7 co t,/d EFPD

. FIGURE 3.1-4 B&W . 3/4 1-32 ' May 2,1975 .

e

e Q

5 3 5 .

,~. . g y 3 g -

5 8 ,7 < a 5 .

4 4 6

> 6 6 7 8 5 5 8 7

, 2. 7. p 3

6 4 6 7 3 3 7 i 1 2 2.

2.

,7 8 5 5 8 7 , 6 4 4 6 ,

6 ..

5 8 7 8 5 .

6 1 2 .6 5 3 5 5

. *l .

y- e FOR ILLUSTRATI0ft O!!L BANK NO. 11005 PRUPOSE.

1 ,

Safety -

2 7 Safety 3 Safety -

4 5- Safety - -

5 12 Regulating

- 6 12 Regulating

7. 8 Regulating -

8 8 APSR CONTROL R0D CORE LOCATION EFPU AND GROU'P ASSIGNMERIS.2 200 t/o

. FIGURE 3.1-5 .

B&W-STS 3/4 1-33 May 2,1975 O

e g 9 g ,

C

t P0;lER DISTRICUTI0ft LIMITS

~

. HOT CHAN!!EL FACTOR - F q f LIMITI?!G CONDITI0ft FOR OPERATION 3.2.2 F qshall be limited by the following relationships:

e. 9 +

I) ~

F q$ p W5

< -( . . -or-,- -

THERMAL POWER 0.

where P = RATED THERMAL POWER and P 1 1

APPLICABILITY: MODES 1 and 2*.

ACTION:

With Fq exceeding its limit: . .

a. within its limit, or

. Immediately bring Fq

b. Immediately reduce THERMAL PO'.iER and the Nuclear Overpower trip setpoint in direct proportion to the excess that Fq exceeds its limit, and .

c. Demonstrate through in-core mapping that n F is within limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY withilf the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE RE0tlIREMENTS 4.2.2 F nshall be determined to be within its limit by using the incore detectors to obtain a power distribution map:

a. Prior to initial operation above 75 percent of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

See Special Test Exception 3.10.3 DAVIS-BESSE, UNIT 1 3/4 2-3 July 3, 1975

s .]

PO'IER DISTRICUTI0il LIMITS

~

l HOT CitAtiNEL FACTOR - F g LIMITIi;G CONDITION FOR OPERATIOil l

N shall be limited by the following relationship: -

3.2.3 F 3H

/ O' '

N F

aH I I' 7#) El + 82(1-P)]

THERMAL POWER where P = RATED THEFJGL POWER and P 1 1 0 APPLICABILITY: MODES 1 and 2*.

ACTION:

N With F 3g exceeding its limit: , ,

- a. Immediately reduce THERMAL POWER in direct proportion to the

- excess that F 3g exceeds its limit and within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months similarly reduce the Nuclear Overpower high trip setpoint.

b. Demonstratethroughln-coremappingthatF H is within limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be ia HOT STANDBY within the next 2 hdurs.

SURVEILLANCE REQUIREMENTS 4.2.3 F 3g shall be determined to be within its' limit by using the incore detectors to obtain a power distribution map:

a. Prior to operation above 75 percent of RATED THERMAL POWER after each fuel loading, and ,

b. At least once per 31 Effective Full Power Days.

See Special Test Exception 3.10.3 DAVIS-5 ESSE, UNIT 1 3/4 2-4 July 3, 1975 l?

)

POUER DISTRIBUTIO?! LIMITS QUADRANT POWER TILT LIMITIf;G C0i:DITIO!! FOR OPERATIO f i 4%

3.2.4 The QUADRA!!T POWER TILT shall not exceed (51%.

1 APPLICABILITY: MODE 1.*

ACTI0il:

a. With the indicated QUADRANT POWER TILT determined to exceed .

J% but _< :(q )%,

1. Imediately correct the power tilt, or

, 2. Reduce THEPJGL POWER so as not to exceed THEPJtAL POWER, including pcwer level cutoff, allowable for the reactor coolant pump combination less two percent for every percent of indicated QUADRAtlT POWER TILT, andin eruss # + N

. 3. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the Nuclear Overpower or the !!uclear Overpower Based on RCS Flow and. AXIAL POWER IMBALANCE trip setpoint 2% for every percent of indicated QUADRANT POWER TILT. in < <ces2 o F Wo

b. Withtheindicated(QUAkRAfiT40Uk ILT determined to exceed M )% or exceeding '{-5)% foi more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , imediately reduce THERIML PO'.lER to (J0)'5 of THEPJML POWER allowable for the reactor coolant pump combination and within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the fluclear Overp]wer trip setpoint to < (J55)5 of O._r THERMAL POWER allowable for the reactor coolant pump combination; be in HOT STANDBY within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLA?;CE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT shall be determined to be within the limits:

a. At least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during operation above 15% of RATED THERMAL POWER except when the QUADRANT POWER TILT alarm is inoperable, thrtr at least once per hour.

6Aen-l l *See Special Test Exception 3.10.1. l DAVIS-BESSE, UNIT 1 3/4 2-5 July 3,1975 IV

_ .. A

8 t I M... ... [ F M 5 Yd T 5 M I $ M N 3 5 8 5 f S 5 M I d F E

w POWER DISTRIBUTION LIMITS _

PROOF & RtVIEW COPY

-,w

- SURVEILLANCE REQUIREMENTS (Continued)

-j.) .

4 ,;

R*.': b. At least once every 31 days by performance of a CHANNEL CHECK between the QUADRANT POWER TILT as measured by the out-of-core g.m.g'dg detectors and the QUADRANT POWER TILT measured by the incore

~+jf]Q detectors, and i,, , ;. g

c. At least once every 92 Effective Full Power Days by performance hr-

' c EP of a CHANNEL CAL.IBRATION of the QUADRANT POWER TILT a IW the incore detectors.

I

~..

_u

4

ra -

?-M I, .&m .

9 V

June 24, 1975 DAVIS-BESSE, UNIT 1 ,

. 3/4 2-6 e

.w..= +9_

PROOF & REVIEW CO:Y TABLE 3.3-1 (Continued)_

TABLE NOTATION 4

With the Reactor Protection System trip breakers in the closed position and the control rod drive system capable of rod withdrawal.

- When Shutdown Bypass is actuates.

amps on both i High voltage to detector may be de-energized above 10-10

- intermediate Range channels:

1820 psig by (a) Trip may be manually bypassed when RCS pressure 1 actuating Shutdown Bypass provided' that:

(1) The Nuclear Overpower trip setpoint is 1 5%'of RATED THERMAL POWER, 1820 The Shutdown Bypass RCS Pressure--High trip setpoint is 1 (2) psig, and The Shutdown Bypass is manually removed when RCS pressure > 19 (3) psig.

ACTION STATEMENTS WiththenumberofchannelsOPEPbSLElessthanrequired ACTION 1 -

by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or open the control rod drive trip breakers.

ACTION 2 - 'With the number of OPERABLE channels one less than th Total !! umber of Channels and with the THERMAL PO l

' a. < 5% of RATED THERMAL POWER, immediately-place-the Inoperable-channel in the tripped condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 55 of RATED THERMAL POWER.

l

b. > 5% of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied:

1,-- The inoperable channel is immediately-placed

.in -the tr-ipped-condi-tion.

4 TheMinimumChannelf0PERABLErequirementis met; however, one edditional channel may be bypassedforupto/jhoursforsurveillance 7 Aa +

testing per Specification 4.3.1.1, ,%

peo,ised

,6e +<,p' J is k lwpnwe etw ,.,,et F A.e r- n r.

W.Cn% cloe,)Y 3/4 3-June 24, 1975 DAVIS-BESSE, UNIT 1 J .

I

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I t

~

PROOi'& rdMEW COeY TABLE 3.3-1 (Continued) l TABLE NOTATION and-the 4noperable channel-above may be-by-passed _for up to-(15) minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per4cd-when--necessary-to-test the~ trip breaker l associated with-the logic of-the-charnel being tested-per-Specification 4.3.1.1, and 2 f. THEPMAL-POWER-is reYtricted-to-<-(75)% of RATED THERMAL-POWER-and-the-Nuclear -Overpower trip -is reduced-to-<-(85)%-of-RATED-THERMAL POWER or the QUADRANT-POWER-TILT 1s monitored at-least once per-8-hours--

ACTION 3 -

With the number of OPERABLE channels one less than the .

Total Number of Channels and with the THERMAL POWER level:

a. < 5% of RATED THERMAL POWER, synediately-place l the-inoperable-channel-in the -tripped-condition;-

restore the lnoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

.b. > 5% of RATED THERMAL POWER, operation may continue l provided both of the following conditions are satisfied:

L - -The-inoperable-channel isJmmediately- placed-in the--tripped condition.

,2 The Minimum Channels OPERABLE requirement is met; however, one additional channel F.y be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveille.ee testing per Specification 4.3.1.1, and ihe-inoperable - -

channel-above may be bypassed for up-to (15) minutes--in-any-24--hour-period,ahen necessary to tes<t_the- trip breaker associated with the logic

.of-the- channel-being-tested-per' Specification ,

4.3rT-1.p h .de d 1& tu ;^ vp ra ble c *" '

tb ; a n,_ t r:ppoc/ C cm/ Mo - dA y v Ars %

Action 4 - With the number 6f channels OPEPABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL Power level:

a. < 5% of RATED THERMAL POWER, restore the inoper- l able channel to OPERABLE status prior to increasing

.TilERMAL POWER above 5% of RATED THERMAL POWER.

DAVIS-BESSE, UNIT 1 3/4 3-4 June 24, 1975

-W

'I l

t l

. PROOF & REV!EW CC TABLE 3.3-1 (Continued)

TABLE NOTATION

b. > 5% of RATED THERMAL POWER, power operation may continue.

! ACTION 5 - Oith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. < 10-10 amps on the Intermediate Range (IR) in-strumentation, restore the inoperable channel to OPERABLE gatus prior to increasing THEPJiAL POWER above 10- amps on the IR instrumentation. ,

b. > 10-10 amps on the IR instrumentation, operation

.- may continue.

ACTION 6 -

With the number of channels OPERABLE one less than re-quired by the Minimum Channels OPERABLE requirement,

~

immediately verify compliance with the SHUTDOWN MARGIN ~

. requirements of Specification 3.1.1.1 and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter.

ACTION 7 - With one reactor trip module fail'ed in the untripped state, either:

a. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;

1. Test the remaining reactor trip modules, and t 2. Restore the failed reactor trip module to

. OPERABILITY and restore porter to the C.1D trip device, or

3. Remove power supplied to the CRD trip device

[ associated with,the failed reactor trip module; ,

or l

b. Be*in HOT SHUID0UN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . l Op u w:~ may e unm u e p m'sded rw +^~

m;& nw e c lo e wis a t*%h le re go 're ement O y c. t . a al +he r e n d .i..~.s tes eder Pr:p n a olk.s ce.g fosfeN ta:! A .% 9 h asrs or be ,s ,% - S H urDcu.1 i

m n. , rha .

n e r +- f- h o u rs.

! DAVIS-BESSE, UNIT 1 3/4 3-5 June 24, 1975

~

+

7 8

f-

5 j

TABLE 3.3-2 E:

G REACTnR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES

& RESPONSE TIMES .

G FUNCTIONAL UNIT N Manual Reactor Trip < O or seconds 1.

d h 2. Nuclear Overpower

i ddJ seconds. .

a

3.

RCS Outlet Temperature--High 1 4 ,</F seconds ,

Nuclear Overpower Based on RCS Flow and*

V*i"J4 & 2 / 3? "'w4- l 4.

AXIAL POWER IMBALANCE G,nf. , //ov i dzJseconds xx

5. RCS Pressure--Low

, i e?>7 seconds ,

g , -

.a. 1 0,#7 seconds

, w 6. RCS Pressure--High a - coa % + h p-m 4,. 1 0 ,*7 seconds

7. Variable Low RCS Pressure .

8. Nuclear Overpower Based on Pump Monitor * ,

1 czJseconds t * ,

T 30

. 9. Reactor Containment Pressure--High

-< o.#/ seconds o

Neutron detectors are exempt from response time testing. ' Response time shall be O

T l

- measured from detector output or input of first electronic component in channel. on kx Z n /y & c 'o n /v R D.S obAry cn. r a' breales- 4/a, ( k u,. A f ,e ,g g ) .

s

.= .

5-0

%e n

O 1, N

k------ .

TS REASON I 3.3.2.1, 4.3.2.1.1, 4.3.2.1.2 Changes reflect that the terminology for DB-1 is the " Safety Features Actuation System (SEAS)

. 4.3.2.1.3 ENGINEERED SAFETY FEATURES RESPONSE TINE is correct terminology in this case since this term is concerned

- with the Actuated Ecuipment and not the Actuating System. .

Table 3.3-3 The DB-1 does not have " Manual Ini-tiation" pushbuttons for the individ-ual components as opposed to other B & W plants that do, i.e., there is no one pushbutton which will start ,

and line up HPI or any of the other functional units. The functional units on page 3/4 3-13 describes all the manual pushbuttons associated with the SFAS. Deletion of the " Manual Ir.itiation" requires deletion of ACTION 8.

Table 3.3-3 Deletion of ACTION 6 requires renumber-

_ ing the other ACTION STATEMENTS.

Tabl,e 3.3-3 Containment Pressure High-High: The ACTION STATEMENT has been changed to new ACTION ,8 as DB-1 cannot bypass just the Containment Pressure High-

- High signal. Therefore the old ACTION 10 statement is the same as the new ACTION 8.

Table 3.3-3 CONTAINMENT ISOLATION: DB-l's SEAS will isolate the containment with each of the trips listed. Each trip will isolate a portion of the pene-trations into containment. Therefore this section cannot be broken down into subsections which would imply complete isolation or complete, purge isolation containment with each signal.

The modes are in agreement with the

NRC's requirements for the FUNCTIONAL UNITS.

l Table 3.3-3 The EMERGENCY HEAT REMOVAL section has i been deleted from this table since the DB-1 SEAS does not do this. Feedwater l isolation, auxiliary feedwater initiation i and main steam isolation is accompliched by the Steam and Feedwater Rupture Con-trol System at DB-1.

3/4 3-27

~

g REASON Table 3.3-3 Manual Actuation Channels were split I for clarity in order to show that there is only one channel for A and one for i

B.

Table 3.3-3 There are 34 Coincidence Logic Channels in the SFAS. Each channal consists of l two output modules and their associated ;

output relays. Both outp,,t u modules and their associated output relays must be trippable for that channel to be operable. Therefore the note and other changes have been made.

ACTION 8, Table 3.3-3 Deleted due to Deletion of manual "

initiation.

ACTION 9 , Table 3.3-3 Renumbered to ACTION 8.. DB-l's SEAS has no interfacing with other control systems. The SEAS therefore meets IEEE-279-1971 with only 3 channels operable. Two (2) hours is not suffi-cient time to run this test. This has been verified at the factory checkout for the ,SFAS.

ACTI,0N (9), Tabla 3.3-3 Deleted; for a three channel system.

ACTION 10, Table 1.3-3 Deleted; see the third comment on Table 3.3-3.

ACIION (10), Table 3.3-3 Deleted' for a three channel system.

ACTION 11, Table 3.3-3 Renumbered to ACTION 9.

New ACTION 10 and 11, Table 3.3-3 Th'ese actions insure that these valves will be in the required position.

ACTION 13 Tripping of the inoperable component in a coincidence logic channel insures that the equipment will be actuated when the operable components 'trfp.

Table 3.3-4 Changes consistent wien Table 3.3-3.

ALLOWABLE VALVES to be supplied by Bechtel.

Table 3.3-5 Changes represent the components actuated by the SEAS. Values to be supplied by Bachtel.

( Table 4.3-2 Changes to be consistent with the above.

i 3/4 3-28

. r .

INSTRIDENTATION

/

3/4.3.2 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I

LIMITING CONDITION FOR OPERATION 1

3.3.2.1 The Safety Features Actuation System instrumentation channels shown in

' Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIME 3 as shown in Table 3.3-5 .

APPLICABILITY: As shown in Table 3.3-3 ACTION:

a. With a Safety Features Actuation System instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values -

column of Table 3.3-4 , either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable and take the ACTION shown in Table 3.3-3 .

b. With a Safety Features Actuation System instrumentation channel inop-erable, take the action shown in Table 3.3-3 .

SURVEILLANCE REQUIREMENTS ,

4.3.2.1.1 Each Safety Features Actuation System instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALI-BRATION and CHANNEL FUNCTIONAL TEST during the modes cnd at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operatio~n. The total bypass function shall be domonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel.affected by bypass operation.

4.3.2.1.3 The ENGINEERED SAFITY FEATURES RESPONSE TIME of each SFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of re-dundant channels in a specific SFAS function as shown in the ' Total No. of Channels" Column of Table 3.3-3.

9 3/4 3-9

- . . . . . - . ~

,s

,j 1

. i.

TABLE.3.3-3 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES

ACTION

1. SAFETY INJECTION

a. High Pressure Injection

1) Containment Pitessure -

High 4 2 3 1,2,3 8

2) RCS Pressure - Low

4 2 3 1,2,3 8

b. Low Pressure Injection M

1) Containment Pressure -

y liigh 4 2 3 1,2,3 8 g 2) RCS Pressure - Low-Low ** 4 2 3 1,2,3 8

2. CONTAINMENT SPRAY

a. Containment Pressure -

High-liigh 4 ,

2 3 1, 2, 3 8 i

3. CONTAINMENT ISOLATION

a. Containment Pressure - High 4 2 3 1, 2, 3, 8

b. Containment Pressure -

liigh-liigh 4 2 3 1,2,3 8

c. Containment Radioactivity -

liigh 4 2 3 ALL MODES 8 and 9

d. RCS Pressure Lov* 5 2 3 1,2,3 8

I l rm _

l' i

1 TABLE 3.3-3 (continued)

SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

FUNCTIONAL UNIT Pr CHANNELS TO TRIP OPERABLE MODES ACTION

3. CONTAINMENT ISOLATION (Conti.4ued)

e. RCS Pressure Low-Low ** 4 2 3 1,2,3 8

4. CONTAINMENT COOLING

a. Containment-Pressure -

High 4 2 3 i

1, 2, 3, 8

. R s-

b. Containment Pressure Hi8h-High 4 2 3 1,2,3 8 Y c. RCS Pressure Low

4 2 3 1,2,3 8 U Atto /HDhAT[R

5. MAIN STEAM 4 SOLATION

a. Containment Pressure - 0 High-High 4 2 3 1,2,3 8

6. CONTAINMENT SUMP SUCTION ,

a. Borated Water Storage Tank -

Low 4 2 3 1,2,3 8 ;

7. DECAY HEAT REMOVAL SYSTEM ISOLATION AND INIERLOCK

a. RCS Pressure 2 1 1 1,2,3 10 I

4

4 i

1 e

/

TABLE 3.3-3. (Continued)

SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM T0rAL NO. CHANNELS CHANNELS

_.PPLICABLE FUNCTIONAL UNIT 2F CHANNELS TO TRIP OPERABLE MODES

ACTION

8. CORE FLOODING TANK ISOLATION VALVE INTERLOCK

a. RCS Pressure 2 2 2 1 11

9. SFAS MANUAL ACTUATION

^( s r w CHANNEL A i 1 1 1 1,2,3,4 12_

w 10. SFAS MANUAL ACTUATION D ACruA T/or/ CHANNEL M 2 1 1 1 1,2,3,4 12 w -

,' 11. CONTAINMENT SPRAY MANUAL

" ACTUATION CHANNEL 1 1 1 1 1,2,3 12

12. CONTAINMENT SPRAY MANUAL ACTUATION CHANNEL 2 1 1 1 1,2,3 12

13. COINCIDENCE LOGIC CHANNELS 34 ,

2*** 2*** All Modes 13

14. SEQUENCE. LOGIC CHANNELS 4 2 3 1,2,3 8 I

l l TABLE 3.3-3 (Continued) ;

i - !

, s, .

TABLE NOTATION i

Trip function may be bypassed in this mode below 1800 psig. Bypass

~

, j shall be automatically removed when RC pressure exceeds 1800 psig.

. ** Trip function may be bypassed in this mode below 600 psig. Bypass shall be automatically removed when RC pressure exceeds 600 psig, i

- - Two output logic modules and associated relays per Coincidence Logic Channel.

ACTION STATEMENTS ACTION 8 -

With the number of OPERABLE channels one less than the Total -

Number of Channels and with RC system pressure:

a. 4 1800 psig, restore the inoperable channel to OPERABLE status prior to increasing the RCS pressure above 1800 psig, t

b. El 1800 psig, operation may continue provided that the minimum channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to six hours for surveillance testing per Specification 4.3.1.2.

ACTION 9 -

With less than the Minimum Channels OPERABLE, operation may i

continue provided the containment purge and exhaust isolation valves are maintained closed.

ACTION 10 - With one or two chahnels inoperable and reactor coolant pressure >280 psig, both Decay Heat Isolation Valves (DH 11 i

! and DH 12) shall be verified closed; With less than the Minimum Channels OPERABLE and reactor coolant pressure 4 280 psig operation of the Decay Heat System may con-tinue; however, both channels shall be OPERABLE prior to increas-ing reactor coolant pressure above 280 psig.

ACTION 11 - With either or both channels inoperable and reactor cc. --

pressure:

(a) 2t ' 800 psig immediately verify the Core Flood Tank Isolation Valves (CF1A and CF1B) are open, (b) 4L 800 psig, remain in Mode 4, 5, or 6 and do not increase reactor coolant pressure above 700 psig until both channels

.are OPERABLE.

-( ACTION 12 - With less than the Minimum Channels OPERABLE, be in COLD SHUR-I DOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

l l

3/4 3-13 e

TABLE 3.3-3 (Continued) lT ACTION STATDTNIS (Continued)

ACTION 13 - With any components in the coincidence logic channels inopera-ble, innediately trip that component or be in COLD SHUIDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

O 9

i e

e D

G f

f 3/4 3-14

.J.

TABLE 3.3-4 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOIlfr ALTR ABLE VALUES .

1. SAFETY INJ1:CTION

a. High Pressure Injection

1. Containment Pressure - High 18.4 psia ( ) psia

2. RCS Pressure - Low 1600 psig ( ) psig

b. Low Pressura Injection

1. Containment Pressure - High 18.4 psia ( ) psia

2. RCS Pressure - Low-Low 400 psig ( ) psig

2. CONTAINMENT SPRAY

~

U a. Containmer.c Pressure - High-High 38.4 psia ( ) psia

3. CONTAINMENT IRJLATION

a. Containment Pressure - High 18.4 psia ( ) psig

b. Containment Pressure - High-High 38.4 psia

( ) psia

c. Containment Radiation - Ifigh 25 mR/hr ( ) mR/hr

d. RCS Pressure Low 1800 psig ( ) psig

e. RCS Pressure Low-Low 400 psig ( ) psig

4. CONTAINMENT COOLING

a. Containment iressure - High 18.4 psia ( ) psia o ' -1

. ,~

c TABLE 3.3-4 (Continued)

N ' ' SYSTEM' I'NSTRUMENT'ATION TRIP SETPOINTS SAFETY' ' FEATURES 'A'C'TUAT'I'O' FUNCTIONAL UNIT TRIP SETPOI E ALLNABLE VALUES .

4. COKIAINMENT COOLING (Continued)

b. Containment Pressure High-High 38.4 psia ( ) psia

c. RCS Pressure Low 1600 psig ( ) psig

5. MAIN STEAM ISOLATION

a. Containment Pressure - High-High 38.4 psia ( ) psia i

6. CONTAINMENT SUMP SUCTION W ,

a. Borated Water Storage Tank - Low

3 feet H2 O ( ) feet H 2O h 7. DECAY HEAT REMOVAL SYSTEM ISOLATION AND INTERLOCK

a. RCS Pressure -

280 psig ( ) psig

8. CORE FLOODING TANK ISOLATION VALVE INTERLOCK .

a. RCS Pressure 800 psig ( ) psig i

4 l

TABLE 3.3-5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual 4

a. Fans

1. Emergency Vent Fan # dE( )*/( )**

2. Containment Cooler Fan # $( )*/( )**

b. HV &AC Isolation Valves

1. ECCS. Room 25 ( )*/( )**

2. Emergency Vettilation 6( )*/( )** ,

3. Containment Air Sample dE( )*/( )**

4. Containment Purge f( )*/( )**

5. Pentration Room Purge $( )*/( .)*k

c. Control Room HV & AC Units f( )*/( )**

d. High Pressure Injection

1. High Pressure Injection Pumps # f( )*/( )**

2. High Pressure Injection Valves ji( )*/( )**

e. Component Cooling Water

1. Component Cooling Water Pumps # 6( )*/( )**

2. Component Cooling Aux. Equip. Inlet Valves $E(, )*/( )**

3. Component Cooling to Air Compressor Valves dE( )*/( )**

f. Service Water System

1. Service Water Pumps # gE( )*/( )**

2. Service Water From Conpone t Cooling 1

Heat Exchanger Isolation Valves f( )*/( )**

g. Containment Spray Isolation Valve s 6( )*/( )**

h. Emergency Diesel Generator Ib ( )

1. Containment Isolation Valves

1. Vacuum Relief f( )*/( )**

2. Normal Sump. JE( )*/( )**

3. RCS Letdown Delay Coil Outlet n( )*/( )**

4. RCS Letdown High Temperature f( )*/( )**

5. Pressurizer Sample 6( )*/( )**

6. Service Water to Cooling Water 6( )*/( )**

7. Vent Header .

6( )*/( )**

8. Drain Tank 6( )*/( )**

9. Core Flood Tank Vent 4( )*/( )**

10. Core Flood Tank Fill f( )*/( )**

11. Steam Generator Sample 6( )*/(, )**

I i

3/4 3-17 l

a TABLE 3.3-5

-. ENGINETRED SAFETY FEATURES ACTUATION SYSTEM RESPONSE TIME INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SEC0 HTS

12. Atmospheric Vent 6( )*/( )**

13. Quench Tank 6( )*/( )**

14., Emergency Sump 6( )*/( )**

15. RCP Seal Return 6( )*/( )**

16. Air Systems f( )*/( )**

17. N2 System $( )*/( )**

18. Quench Tank Sample 6( )*/( )**

19. Main Steam Warmup Drain 6( )*/( )**

20. Makeup s( )*/( )**

21. RCP Seal Inlet f( )*/( )**

22. Core Flood Tank Sample 6( )*/( )** ~

23. RCP Standpipe Demin Water Supply f( )*/( )**

24. Containment H Dilution Inlet ${ )*/( )**

25. Containment H Dilution Outlet 6( )*/( )**

j. BWST Outlet Valves 6( )*/( )**

k. Low Pressure Injection

1. Decay Heat Pumps # 6( )*/( )**

2. Low Pressure Injection Valve s 6( )*/( )**

3. Decay Heat Pump Suction Valves' 6( )*/( )**

4. Decay Heat Cooler Outlet Valves d( )*/( )**

5. Decay Heat Cooler Bypass Valves f( )*/( )**

1. Containment Spray Pump # 6( )*/( )**

m. Component Cooling Isolation Valves

- 1. Inigt to Containment 5( )*/( )**'

2. Outlet from Containment f( )*/( )**

3. Inlet to CRDM's d( )*/( )**

4. CRDM Booster Pump Suction d( )*/( )**

n. Steam and Feedwater Isolation Valves

)*/(

1. Main Steam Line 6( )**

2. AFPT Inlet 5( )*/( )**

3. Main Feedwater Stop d( )*/( )**

4. Auxiliary Feedwater 6( )*/( )**

5. Main Steam Warmup 6( )*/( )**

6. AFFT Alternate Inlet 6( )*/( )**

e i

l-3/4 3-18

TABLE 3.3-5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM RESPONSE TIME e INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i 2. Containment Pressure - High

a. Fans

1. Emergency Vent Fan # 6( )*/( )**

2. Containment Cooler Fan # 6( )*/( )**

b. HV & AC Isolation Valves

1. ECCS Room 6( )*/( )** *

2. Emergency Ventilation 6( )*/( )**

3. Containment Air Sample 6( )*/( )**

4. Containment Purge g( )*/( .)**

5. Penetration Room Purge 6( )*/( )**

c. Control Room HV & AC Units 6( )*/( )**

d. High Pressure Injection

1. High Pressure Injection Pumps # , 6( )*/( )**

2. High Pressure injection Valves 6( )*/( )**

e. Component Cooling Water ,

1. Component Cooling Water Pumps # 6(' )*/( )** ,

2. Component Cooling Aux. Equip. Inlet Valvesn( )*/( )**

3. Component Cooling to Air Compressor Valves 6( )*/( )**

f. Service Water System

1. Service dater Pumps # 6( )*/( )**

2. Service Water From Component Cooling Heat Exchanger Isolation Valves b( )*/( )**

g. Containment Spray Isolation Valve s 6( )*/( )**

h. Emergency Diesel Generator 6( )*/( )**

1. Containment Isolation Valves

1. Vacuum Relief 6( )*/( '**

2. Normal Sump f( )*/( )**

3. RCS Letdown Delay Coil Outlet 6( )*/( )**

4. RCS Letdown High Temperature g( )*/( )**

5. Pressurizer Sample g( )*/( )**

3/4 3-l') ,

TABLE 3.3-5 ENCINEERED S M TV FEATURES ACTUATION SYSTEM RESPONSE TIME INITIATING SIGNAL AND FUNCTIONS RESPONSE TIME IN SECONDS

6. Service Water to Cooling Water f: ( )*/( )**

7. Vent Header f- ( )*/( )**

8. Drain Tank 6( )*/( )**

9. Core Flood Tank Vent dE( )*/( )**

10. Core Flood Tank Fill gE( )*/( )**

11. Steam Generator Sample 6( )*/( )**

12. Atmospheric Vent f( )t/( )**

13. Quench Tank 6( )*/( )**

14. Emergency Sump 6( )*/( )**

15. RCP Seal Return 6( )*/( )**

16. Air System 6( )*/( )**

. 17. N 2 System 6(, )*/( )** .

18. Quench Tank Sample 6( )*/( )**

19. Main Steam Warmup Drain 6( )*/( )**

20. Makeup gb( )*/( )**

21. RCP Seal Inlet s( )*/( )**

22. Core Flood Tank Sample gE( )*/( )**

23. RCP Standpipe Demin Water Supply fE( )*/( )**

j 24. Containment H2 Dilution Inlet IS( )*/( )**

25. Containment H2 Dilution Outlet 6( )*/( )** .

6( )*/( )**

j. BWST Outlet Valves

k. Low Pressure Injection

1. Decay Heat Pumps # 6( )*/( )**

2. Low Pressure Injection Valves es( )*/( )**

3. Decay Heat Pump Suction Valves tE( )*/( )t*

4. Decay Heat Cooler Outlet Valves eg( )*/( )**

5. Decay Heat Cooler Bypass Valves gg( )*/( )**

l 3. Containment Pressure - High High

a. d'( )*/( )**

Containment Spray Pump # :

b. Component Cooling Isolation Valves

1. Inlet to Containment 6( )*/( )**

2. Outlet from Containment 6( )*/( )**

3. Inlet to CRDM's 6( )*/( )**

4. CRDM Booster Pump Suction (E( )*/( )**

c. Steam and Feedwater Isolation Valves

1. Main Steam Line gg( )*/( )**

s 3/4 3-20 m

TABLE 3.3-5

( ENGINEERED SAFETY FEATURES ACTUATION SYSTEM RESPONSE TIME INITIATING SIGNAL AND FUNCTIONS RESPONSE TIME IN SECONDS 9

2. AFPT Inlet f$ ( )*/( )**

3. Main Feedwater Stop at )*/( )**

4. Auxiliary Feedwater 3[ (( )*/( )**

5. Main Steam Warmup )*/( )**

6. AFPT Alternate Inlet ][((

gg )*/( )#*

4. Reactor Coolant Pressure Low

a. Fans

1. Emergency Vent Fan # d5( )*/( )**

2. Containment Cooler Fan # (=( )*/( )**

b. HV & AC Isolation Valves

1. ECCS Room g$( )*/( )**

2. Emergency Ventilation $( )*/( )**

3. Containment Air Sample gE( )*/( )**

4. Containment Purge :h( )*/( )**

5. Penetration Room Purge ,

db( )*/( )**

,c. Control Room HV & AC Units 26 ( )*/( )**

d. High Pressure Injection

1. High Pressure Injection Pumps # d'(

)*/( )**

2. High Pressure Injection Valves f6( )*/( )**

e. Component Cooling Water Pumps # !!( )*/( )**

f. Service Water System

1. Service Water Pumps # fE( )*/( )**

2. Service Water from Component Cooling Heat Exchanger Isolation Valves d!( )*/( )**

I

g. Containment Spray Isolation Valve s a dh( )*/( )**

h. Emergency Diesel Generator' I;( )

i. Containment Isolation Valves

1. Vacnum Relief 4. )*/( )**

2. Normal Sump 2[(( )*/( )**

3. RCS Letdown Delay Coil Outlet "[ )*/( )**

.4. RCS Letdown High Temperature ][(( )*/( )**

5. Pressurizer Sample

~

f( )*/( )**

3/4 3-21

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

TABLE 3.3-5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM RESPONSE TIME INITIATING SIGNAL AND FUNCTIONS RESPONSE TIME IN SECONDS

6. Service Water to cooling Water db( )*/( )**

7.. Vent Header dh( )*/( )**

8. Drain Tank gb( )*/( )**

9. Core Flood Tank Vent dE( )*/( )**

10. Core Flood Tank Fill jf( )*/( )**

11. Steam Generator Sample gb( )*/( )**

12. Atmospheric Vent gg( )*/( )**

13. Quench Tank gg( )*/( )**

14. Emergency Sump jc( )*/( )**

15. RCP Seal Return dh( )*/( )**

". 16. Air Systems 3;( )*/( )** -

17. N2 System gb( )*/( )**

18. Quench Tank Sample g6( )*/( )**

19. Main Steam Warmup Drain gS( )*/( )**

20. Makeup gc( )*/( )**

21. RCP Seal Inlet gg( )*/( )**

22. Core Flood Tank Sample 4( )*/( )**

23. RCP Standpipe Demin Water Supply gL ( )*/( )**

24. Containment H Dilution Inlet gg( )*/( )**

25. Containment H Dilution Outlet gg( )*/( )**

j. BWST Outlet Valves . gb( )*/( )**

5. Reactor Coolant System Pressure Low-Low

a. Low Pressure Injection

1. Decay Heat Pumps # dk( )*/( )**

2. Low Pressure Injection Valves db( )*/( )**

3. Decay Heat Pump Suction Valves !b( )*/( )**.

4. Decay Heat Cooler Outlet Valves Jf( )*/( )**

5. Decay Heat Cooler Bypass Valves th( )*/( )**

b'. Component Cooling Isolation Valves

1. Auxiliary Equipment Inlet AE( )*/( )**

2. Inlet to Air Compressor- f.,( )*/( )**

6. Containment Radiation - High

a. Emergency Vent Fan # dh( )*/( )**

b. HV & AC Isolation Valves

1. ECCS Room g( )*/( )**

l l

3/4 3-22 i l

d

TABLE 3.3-5 ENGINEERED SAFETY FEATURES ACTUATION SYSIEM RESPONSE TIME INITIATING SIGNAL AND FUNCTIONS RESPONSE TIME IN SECONDS

2. Emergency Ventilation dh( )*/( )**

3., Containment Air Sample dE ( )*/( )**

4. Containment Purge f$( )*/( )**

5. Penetration Room Purge dh( )*/( )**

c. Control Room HV & AC Units jb( )*/( )**

, 7. Borated Water Storage Tank Low

a. Contai,nment Sump Suction Valves 6-( )*/( )** .

b. BWST Outlet Valve s j$( )*/( )**

TABLE NOTATION

Diesel generator starting and sequence loading delays included.

- Diesel generator starting and sequence loading delays not included.

Response time limit includes movement of valves and attainment of pumps or blower discharge pressure. ,

0 4

1 3/4 3-23

o TABLE 4.3,2 SAFETY FEATURE ACTUATION SYS1LM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT CHANNEL CllANNEL OIANNEL MODES IN MIIQI CllECK CALIBRATION FUNCTIONAL SURVEILLANCE TEST REQUIRED ,

I I

i

1. SAFETY INJECTION !

a. High Pressure Injection

1) Containment Pressure-High S R M 1,2,3

2) RCS Pressure-Low S R M 1, 2, 3 ,

i

b. Low Pressure Iajection

. M l

[ 1) Containment Pressure-High S R H 1,2,3 l

,i, 2) RCS Pressure--Low-Low S R H 1, 2, 3 I

2. CONIAINMENT SPRAY al Containment Pressure-- S R M 1,2,3 liigh-liigh

3. CONTAINMENT ISOLATION ,

a. Containment Pressure-High S R M 1, 2, 3, 4, 6

b. Containment Pressure-- S R H . 1,2,3 High-liigh

c. Containment Radioactivity- S R H ALL 190 DES liikh

d. RCS Pressure Low S R M 1,2,3

e. RCS Pressure -- Low-Low S R. M 1, 2, 3

.i _

l TABLE 4.3-2 (Continued) .

SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CIIANNEL MODES IN WilICH CllANNEL CllANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED

4. CONTAINMENT COOLING ,

a. Containment Pressure-II,igh S R M 1,2,3,4

b. Containment Pressure--

liigh-liigh S R M 1,2,3

c. RCS Pressure- Low S R M 1,2,3

5. MAIN STEAM ISOLATION R

A, Containment Pressure --

H 1,2,3 liigh-1,ligh S R Y

U 6. CONTAINMENT SUMP SUCTION Borated Water Storage Tank-Low S R . M 1,2,3

7. DECAY llEAT REMOVAL SYSTEM ,

ISOLATION AND INTERLOCK

a. RCS Pressure S R M 1,2,3

8. CORE FLOODING TANK ISOLATION VALVE INTERLOCK

a. RCS Pressure S R M 1

9. SFAS MANUAL ACTUATION CHANNEL A N.A. N.A. R 1,2,3,4

10. SFAS MANUAL ACTUATION CilANNEL B ,N.A. N.A. R 1, 2, 3, 4

11. CONTAINMENT SPRAY ACTUATION CIIANNEL 1 N.A. NA.A R 1, 2, 3

TABLE 4.3-2 (Continued) : -

SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CIIANNEL MODES IN WHICH l CilANNEL CllANNEL FUNCTIONAL SURVEILLANCE !

-FUNTIONAL UNIT CllECK CALIBRATION TEST REQUIRED

12. CONTAINMENT SPRAY ACTUATION N.A. N.A. R 1,2,3 CllANNEL 2 1

13. COINCIDENCE LOGIC CHANNELS N.A. N.A. M ALL MODES ]

14. SEQUENCE LOGIC CilANNELS N.A. N.A. M 1,2,3 l

kt.

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,d IllSTRUMENTATI0n PROOF & u.cy'IEW COrY 3/4.3.3 MONITORING INSTRUMENTATION

~ ,e4 RADIATION MONITORING

'DM -

yK LIMITING CONDITION FOR OPERATION YMM

.f.,. ?)

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3.3.3.1 The radiation monitoring instrumentation channels shown in Table (3.3-6) shall be OPERABLE with their alarm / trip setpoints within the

7ir- - specified limits.

, APPLICABILITY: As shown.in Table (3.3-6).

ACTION:

kW With a radiation monitoring channel alarm / trip setpoint M a.

exceeding the value shown in Table ( .3-6), adjust the setpoint to within t.he limit within 2jhours or declare the channel inoperable. ggG)g, b.

~

With one or more radiation monitoring) channels inoperable,The h '

take the ACTI0ft shown in Table (3.3-6 .

Specifications 3.0.3 and 3.0.4 are not applicable.

-. j

_N

SURVEILLANCE REQUIREMENTS Each radiation monitoring instrumentation channel shall be I: f".' 4.3.3.1 demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL

'% CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes at Mni '

the frequencies shown in Table (4.3-3).

se L

March 20, 1975 DAVIS-BESSE, UNIT 1 ,

3/4 3-26

--d. l 8 l *

~- .. .

TABI'E 3.3-6 -

[

o RADIATION MONITORING INSTRUMENTATION iE % k ,

y MINIMUM APPLICABLE ALARM / TRIP MEASUREMENT.

cn CHANNELS SETPOINT RANGE ACTION O INSTRUMENT '

OPERABLE MODES M

~ 1. AREA MONITORS !

c %.24 N4 4// h 2 -

5 a. Fuel Storage = Pee 1 Area H i. 6riticalii.y Munii.ur (1)

1 15 mR/hr (10'j - 10 4) mR/hr 13 ' ;

' . +k--Vat 11attorrSystem f I

- ~~< 2 x background (1 - 10') . cpm, 1fr l Istrtatinn (1) b Spem r Fvel /)rea . pgg* l

b. Containment /1rca ma "a r i ir---Purge-&-Exhaust olorn f i (1) 6 1 2 x background (1 ) cpm- 16 ,;

IJnlation

. mR/g,

2. PROCESS M0FTORS

[ a. Fuel Storage Pool Area

1. Gaseous Activity -

U i

Aform./!entilation System

"'#. M,M<. ## ~ ## .c

- 1 2 ^ background- ) cpm 15

. Isolation -(l) (1

~

11. Particulate Activity -

p.m.Nentilation System

/"'# r 12Mackgroun/cc..

N - /d Sc (I M O .) cpm

$ 2 Isolation (1) ** d-0 15 C b.' Containment O (

i. Gaseous Activity 6 n

'# ' '# ~1 W M Purge & Exhaust dIi Juo ~ ' fund. ) cpm b '6 Isolation (1) 6 1

E-:t-bacho/_ (1]-lbQpm '

(1 9 p RCS Leakage Detection (1) l', 2, 3, & 4 1 X x background .

p; '6

'# r 5 ii. Particulate Activity e b, }

g sva ~tPurge & Exhaust /4 / o % /u,

/o-'Og Q

- : Isolation (1) "w6 d**

- < 2 x backgrourid-- (l yl0c) cpm FO 16 U

g RCS Leakage Detection (1) 1, 2, 3 & 4

< A' x background (1 - 10$ cpm Q 14 M /. / Ta 19 (')

L

With fuel in the storage pool or building {e ,

- With irradiated fuel in the storage pool . g.

- L- - - - - - _

"F 00<-;g Puv s n J L=x 4.m f

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.' IgTRUMENTATION_ ..

INCORE DETECTORS ,

.'$ ,k LIMITING CONDITION FOR OPERATION '

W]

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3.3.3.2 As a minimum, the incore detectors shall be OPERABLE as speci-r

'Q' g~ig fied below.

Q.,.:i;,4

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a. For AXIAL POWER IMBALANCE measurements:

.Y$

1. Three detectors, one in each of 3 strings, shall lie in s/ the same axial plans with 1 plane in each axial core half.

d. ~.! . .

2. The axial planes in each core half.shall be symmetrical ~

about the core mid-plane.

3. The detector strings shall not have radial symmetry.

t 5 For QUADRANT POWER TILT measurements:

1. .Two sets of 4 detectors shall lie in each core half.

Each set of detectors shall lie'in the same, axial plane.

V The two sets in the same core half may lie in the same

)

axial plane.

"3 - 2. Detectors in the same plane shall have quarter core radial symmetry.

APPLICABILITY: When the incore detection system is used for serveillance of: .

EM

. - -- a. The AXIAL POWER BALANCE, or

b. The QUADRANT POWER TILT. -

l

~m f ACTION: -

I With less than the specified minimum incore detector arrangement OPERABLE, do not use incore detector measur.ements to determine AXIAL POWER IMB or QUADRANT POWER TILT.

- SURVEILLANCE REQUIREMENTS 4.3.3.2 The incore detector system shall be demonstrated OPERABLE:

June 24, 1975 DAVIS-BESSE, UNIT 1 3/4 3-30

) .

- - + - . .

g Reason

'([ 3.3.3.3 &

4.3.3.3 Delete the parenthesis as these are the applicable Tables.

Table 3.3-7 The instrument names have been changed to confirm to DB-1 terminology. The range of the trigger cannot be specified,

..: therefore, the frequency response has been specified. The seismic trigFer is indicated in the cabinet room, directly behind the control panels.

O 9

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~ . ___ ..____ _ _ __..._ _ _ .

INSTRUMENTATI0ft l SEISMIC I!!STRUMENTATION q

l i LIMITING CONDITI0ft FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table (3.3-7) shall be OPERABLE. ,

APPLICABILITY: ALL MODES.

ACTION:

. a. With the number of OPERABLE seismic monitoring instruments less than required by Table (3.3-7); restore the inoperable instrument (s) to OPERABLE status within 30 days. The pro-visions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. With one or more seismic monitoring instruments incperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the pl.ans for restoring the instrument (s) to OPERAELE status.

~

~ '

SURVEILLAf:CE REOUIREMENTS .

4.3.3.3.1 Each' of the above seismic monitoring instrumentations shall be demonstrated OPERABLE by the performance of the CHAtlNEL CHECK, CHANNEL CALI3 RATIO t and CHANNEL FUNCTIONAL TEST operations at the frequencies showa in Table 14.3-4 5 4.3.3.3.2 Each of the above seismic monitoring channels actuated during a seismic event shall be restored to OPERABLE status and a CHANNEL CALIBRATION performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the seismic event. Data shall be retrieved from actuated channels and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be pre-pared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety. .

\,_

o DAVIS-BESSE, UNIT 1 , 3/4 3-32 -

July 3, 197S.

.q

TABLE 4.3-4 1

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

( '"

' CHANNEL

CHANNEL CHANNEL FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

! 1. Strong Motion Triaxial Accelerometers

a. Containment Concrete Foundation @Elev. 565 M* R SA

b. Containment Interior Secondary Shield Wall @ Elev. 653 M* R -

SA

c. Auxiliary Building Basement Floor

@ Elev. 545 M* R SA

d. Station Site - 300 ft from closest station structure M* R SA -

2. Peak Recording Accelerometers. I

a. Shield Building Top @ Elev. 809 NA R~

~

NA

b. Auxiliary Building Roof @ Elev 660 NA R NA

c. Control Room @ Elev. 623 NA R NA

3. Seismic Trigger .

a. Station Site - 300 ft from closest station **M R SA structure

E scept seismic trigger '

- With cabinet room indication 4

~.4 7

DAVIS-BESSE, UNIT 1 3/4 3-34 June 4, 1975 f

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[ TABLE 3.3-7 I SEISMIC MONITORING INSTRUMENTATION L

(^-. MINIMUx MEASUREMENT INSTRUMENT INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

~1. Strong motion triaxial accelerometers

a. Containment Concrete Foundation '

@ Elev. 565 i 18 1

b. Containment Interior Secondary

' Shield Wall @ Elev. 653 i 8 1 ,

l

c. Auxiliary Building Basement Floor j

@ Elev. 545 --. yg, 1 l

d. Station site - 300 ft from closest , y q i station structure. '

1

! 2. Peak Recording Accelerometers

! a. Shield Building Top @ Elev. 809 il8 1 !

b. Auxiliary Building Roof @ Elev. 660 i18 L

c. Control Room @ Elev. 623 i18 1

+

3: Seismic Trigger

-t ,

t .

,! s. Station site - 300 ft from closest station structure 0.053-20Hz* 1**

Frequency Response

- With cabinet r'oom indication l'

LDAVIS-BESSE, UN"IT 1 - 3/4 3-33 June 4, 1975

~

_--.w q , s .-p , e - =<-e ,-

. TABLE 3.3-9 '!

b REMOTE SHUTDOWN MONITORING INSTRUMENTATION l-T' MINIMUM lE READ 0UT MEASUREMENT CHANNELS .

INSTRUMENT ; LOCATION RANGE OPERABLE c 1. Reactor Trip Breaker Indication (a) 480v F&DC CH. 2 OPEN-CLOSE (a) 1 (Primary trip breaker q switchgear room Unit A)

(b)48CvE&DCCH.1 (b) 1 (Primary trip breaker Switchgear Room UnitB) .

2. Reactor Coolant Temperature . Aux. Shutdown Panel 520-620 F 1 Hot Leg

3. Reactor Coolant System Pressure Aux. Shutdown Panel 0-2500*ggs;2 1 .

4. Pressurizer Level Aux. Shutdown Panel 0-320 inches 1

[ 8 O 5. Steam Generator Outlet Steam Aux. Shutdown Panel 0-1200 psig 1/ steam generator Pressure

6. Steam Generator Startup Range. Aux. Shutdown Panel 0-250 inches 1/ steam generator i

7. Control Rod Position Control Rod Drive 0, 25, 50, 75 1 (per rod)

Limit Switches Control Ccbinets, and 100%

System Logic .

p Cabinet # 4 iS 3 - -

.y O

e mme f

TS REASON 3.3.3.6 Delete the "S" in the title as there is only one system.

3.3.3.6 There are two chlorine detections located in one control room ventilation air intake. Therefore, change this

, spec in order to prevent confusion.

t B3/4.3.3.5 Change " SHUTDOWN" to " STANDBY" since the Remote Shutdown Instrumentation is designed to shut the plant down to the old definition of HOT SHUTDOWN i.e., Tave 2 5250F. With

. the NRC's present definition of HOT SHUTDOWN i.e., 2800F

>Tave > 200 F, the Remote Shutdown Instrumentation can only take the plant to HW STANDBY. This again points out the inconsistency of the NRC's definition of HOT SHUTDOWN and the industry's interpretation of HOT SHUT-DOWN in 10CFR50. .

B3/4.3.3.6 Delete the "S" from " systems" - see comment on 3.3.3.6 Revise the spec as shown. Upon a high chlorine concentra-tion, the normal control room ventilation system is autoatatically shutdown and isolated. The operator may then manually initiate the control room emergency ventil-ation system in the recirculator mode.

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CHLORINE DETECTION SYSTEM 3'

.; LIMITING CONDITION FOR OPERATION .

).$Q 1

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Nffy 3.3.3.6 The chlorine detection system, with the a arm / trip setpoint 8 adjusted to actuate at a chlorine concentration of NTppm, shall be

,U .$* OPERABLE with, as a minimum, (two) OPERABLE chlorinTdetectors located-atX w ,,j , _ .wn m,

,., g

M ,

. a. Reactor _. control-room ventihtion-air- int'ake.

3k3C

-j Mear tnr- control-room-ventilation--air -intake:

APPLICABILITY: ALL MODES ACTION: -

a. With one chlorine detector inoperable, restore the inoperable detector to OPERABLE status within 7' days or within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months initiate and maintain operation of the control room emerger.cy ventilatio'n system' in the recirculation mode' of Q operation until the inoperable chlorine de+2ctor is restored I

to OPERABLE status.

' ~ ' b. . With the chlorine detection system inoperable, initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation until the chlorine

' detection system with at least two OPERABLE detectors is restored to OPERABLE status.

< 20 percent were revealed during previous inspections. .

2. Tubes in those areas where design and experience have '
indicated potential problems.
e O
DAVIS-BESSE, UNIT 1 3/4 4-5 May 19, 1975
'C
e5 REACTOR COOLANT SYSTEM . .
. SPECIFIC ACTIVITY
! LIMITING CONDITION FOR OPERATION
~
< 3.4.8 ' The specific activity of the primary coolant shall be limited to:
I a. <.1.0 pCi/ gram DOSE EQUIVALENT I-131, and l ,
b. < 100/E pCi/ gram.
APPLICABILITY: MODES 1, 2, 3, 4 and 5. .
ACTION: ge& 9' .
,94 :
- MODES 1, 2 'and 3*: .
a. With the specific activity of the p,rimary coolant > 1.0 -
pCi/ gram DOSE EQUIVALENT I-131 but.within the allowable limit. -
t (below and to the left of the linekshown on Figure 3.4-1, /
l ' operation may continue for up to d8/ hours provided that /
l operation under these circumstances shall not exceed 10% of.
the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.
7 _
.b. 'With the specific activity of the primary coo A 1 nt > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48> hours during
. 'one continuo'us time interval or exceeding the imit line shown on Figure 3.4-1,'be in HOT STANDBY with T,yg < 550 F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
c. With the specific activity of the primary coolant > 100/E
, pCi/ gram, be in HOT STANDBY with T,yg < 550*F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
MODES 1, 2, 3, 4 and 5:
a. With the specific activity of the primary _ coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4 until the specific activity of the primary coolant is restored to within its limits; prepare and submit to the l
Commission an ABNORMAL OCCURRENCE Report pursuant to
.. Specification 6.9.1. In addition to the information required by the ABNCRMAL OCCURRENCE Report, this report shall contain the results of the specific activity analyses together with the following information:
l-
WithT,yg>_(J00)*F. - '
aJ6 DAVIS-BESSE, UNIT 1 3/4 4-13 .
July 23, 1975 me i
f
~
4 TABLE 4.4-2 i PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM
. . TYPE OF MEASUREMENT AND ANALYSIS .
MINIMUM FREQUENCY
1. Gross Activity Determination 3 times every 7 days with a maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples ,
2. Isotopic Analysis for DOSE 1 per 14 days EQUIVALENT I-131 Concentration
3. Radiochemical for if Determination 1 per 6 months
4. Isotopic Analysis for Iodine .
. a). Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Includir.g I-131. I-133, and I-135 . the DOSE EQUIVALENT I-131 -
. exceeds 1.0 pCi/ gram, and One sample between 2 and 6
(}{E hours following a THERMAL POWEI
[ ~ change exceeding 15 percent
. of the RATED THERMAL POWER within a one hour period.
l .
l J
j DAVIS-BESSE, UNIT I 3/4 4-15 July 3, 1975
~
t]
REACTOR C00LAriT SYSTEM 3/4.4.10 STRUCTURAL IflTEGRITY .
QUALITY GROUP A COMPONENTS LIMITINGC0::0ITI0ftFOROPERATIdN 3.4.10,1.The structural integrity of components, except steam generator tubes, identified in Section -tG.2.?fof the FSAR as Quality Group A
. components shall be maintained consistent with the acceptance criteria in Specification 4.4.10.1.
APPLICABILITY: MODES 1, 2, 3 and 4. .
ACTION: -
a. With the structural integrity of any of the above components not conforming to the above requirements and Tav 200 F,
, either immediately isolate the affected componen$ o>r be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ,
b. With the structural integrity of any of the above components' .
not conforming to the above requirements and the unit in COLD SHUTDOWN, restore the structural integrity of the affected component to within its limits prior to increasing the Reactor Coolant System temperature above the minimum temperature required by NDT considerations.
SUPVEILLANCE REQUIREMENTS _
4.4.10.1 The following inspection program shall be performed:
a. Inservice Inspections The structural integrity of the Quality Group A components shall be demonstrated by verifying their acceptability per the requirements of Articles IS200 and IS500 of Section XI of the ASME Boiler and Pressure Vessel Code, dated 1971, including Summer and Winter Addenda, as outlined by the inspection program shown in Table 4.4-4.
An initial report of any abnormal degradation of the structural integrity of the Quality Group ~ A components detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.
DAVIS-BESSE, UNIT 1 3/4 4-22 -
July 3,,1975 l
![
1 REACTOR COOLANT SYSTEf4 -
.O ,
a - -SURVEILLA!!CE' REQUIRE!4ENTS (Continued) '
The Inservice Inspection Program shall be reviewed every 5 I
years to assure that'the equipment, techniques and procedures being utilized are current and applicable. The results of these reviews shall be reported in Special Reports to the Comission pursuant to Specification 6.9.2 within 90 days of completion.
b. Inspections Following Repairs or Replacements The structural integrity of the reactor coolant system shall-be demonstrated after completion of all repairs and/or replacements to the system by verifying the repairs and/or replacements meet the .
requirements of Articles IS-400 and 15-500 of Section XI of
, the ASME Boile'r' and Pressure Vessel Code, dated 1971, includ~.ig l
Summer and Winter Addenda. When repairs and/or replacements are made which involve new strength-welds'on components greater o
than 2 inch diameter, the new welds shall receive a surface and 100 percent volumetric examination and meet applicable code requirements. When repairs and/or replacements are made which involve new strength welds on components 2 inch diameter or smaller, the new welds shall. receive a surface examination -
-and meet applicable code requ'irements.
.c. . Inspections Followino System Opening The structural integrity
, of the reactor coolant system sha,11 be demonstrated after each
~ ' closing by performing a leak test, with the system pressurized
' to at least 1.b g' tT5ecTiiid'J22Wr) XI of the psig, ASME in accordance Boiler and Pressurewith Article Vessel Code, IS-500 dated 1971, including Summer and Win'ter Addenda, and the pressure-temperature limits of Specification 3.4.9.1.
d. Pipe Snubber Insoections The OPERABILITY of the hydraulic
, pipe snubbers shall be demonstrated initially at least once after not less than 4 months or.more than 6 months of operation-and then at least once per 18 months by verifying that the snubber hydraulic fluid reservoirs are filled to between the <
minimum and maximum level indication marks.
. e. Internals Vent Valves Each internals vent valve shall be demonstrated OPERABLE at least once every 18 months by visual j inspection and manual actuation.
f. Reactor Coolant Pumo Flywheels The structural integrity of each 4 reactor coolant pump flywheel shall be demonstrated in accordance with the provisions of flRC Regulatory Guide 1.14, Reactor Coolant
- Pump Flywheel Integrity.
-9 DAVIS-BESSE, UNIT l! 3/4.4-23 June 4,1975 e
As stated on page 3.3-3 of FSAR Chapter 16 states, "The boroa concentration is based on the amount of boron required to maintain the core 1 percent subcritical r"N at 700 F. without any control rods in the core. This concentration is 1585 ppm boron while the minimum value specified in the tanks is 1800 ppm boron." Therefore, 1585 ppm is the limit which must be considered when specifying the appropriate number for Surveillance Requirement 4.5.1 b. TEco has proposed that 0.5 ft be used for this number. 0.5 ft is the smallest increment on the operator's level i indication in the control room and will satisfy the 1585 ppm criteria as shown below. ,
1 Assume: 1. The Core Flood Tank is at minimum volume of 7,555 gals.
2. The boron concentration is at minimum -mncentration of 1800' ppm.
, 3. 0.5 ft of demineralized water is added t o the tank.
a. Conversion of 0.5 ft level change to gallons Core Flood Tanks Inside Diameter = 111.625" I.D. (from B&W Core Flood Tank Instruction Manual).
V =gf d 4 2/Y\ [')[
2' V= TI (111.625")2 1 ft (0.5 ft) (144"2j 4
u V = 33.98 ft3 -
V = 33.98 ft 3 x gal
.1337 ft 3 V = 254.15 gal.
Therefore, a level increase of 0.5 ft with 7,555 gals in the Core Flood Tank corresponds to an addition of 254.15 gal. The volume in the Core Flood Tank is now 7809.15 gallons.
b. Resultant boron concentration after addition of 254.15 gala of demineralized. water.
(7555 gals) (1800 ppm) = (7809.15 gals) (X)
X = 1741.42 ppm
c. Results: An increase of 0.5 ft level in the Core Flood Tank, ass 2 ming this increase to be deminer sliced water, will result in a final baron l
concentration of 1741.42 ppm. This meets the 1585 ppm criteria. The l
attached specs have been revised to reflect this.
4
_._,._j_ ,
' * ~d
. l' (]
p 3/4.5 EMERGEflCY CORE C00LIllG SYSTEMS (ECCS)
CORE FLOODIflG TANKS LIMITING C0f!DITI0ft FOR OPERATI0ft 3.5.1 Each reactor coolant system core flooding tank shall be OPERABLE withi
. a. The isolation valve open.and the power to the valve operator is disconnected.49 7emoval of the breakeFTr5HFthe7iretrit;- l Yk
b. bet een 7,555 and 8,004. ga11ons of borated water, .
c. A minimum boron concentration of 1800 PPM, and
d. A nitrogen cover-pressure of between 575 and 625 psig.
APPLICABILITY: MODES 1, 2 and 3*.
ACTION: -
a.. With one core floodir.g tan!: inopera'bic, cxcept as a ra'sult of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
b. With any core flooding tank inoperable due to the isolation valve being closed, either immediate.ly open the isolation valve or be in HOT STAliDBY within one hour and be in HOT SHUTDOWil within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
SURVEILLAtlCE REQ'.IREMEflTS 4.5.1 Each core flooding tank shall be demonstrated OPERABLE:
. a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1. Verifying the water level and nitrogen cover-pressure in the tanks, and
2. Verifying that each tank isolation valve is open.
l With Reactor Coolant pressure > 800 psig. -
~
DAVIS-BESSE, UNIT 1 3/4'5-1 July 3 1975 I9
- EMERGENCY CORE COOLING SYSTEMS
(-
SURVEILLANCE REQUIREMENTS (Continued) y lan~cl itvel
b. At least once per 31 days and at eachtselutimmlume~ increase of > ('1% ci_wak;veh:=e)'by verifying the boron concentration of the tank solution. , p. f .
c. At least once per 31 days by verifying that power to the isola-tion valve operator is disconnected by removing the breaker from the circuit.
d. Demonstrating check valve operation at least once per 18 l .
months.
e. Verifying at least once per 18 months that the core ' flooding tank isolation valves open automatically and are interlocked against closing whenever the Reactor Coolant System pressure exceeds 800 psi @}
/
e e
g.
e D
e e
e l DAVIS-BESSE, UNIT 1 3/4 5-2 July 3, 1975 L
~ .
9 EMERGEflCY CORE C00LIf!G SYSTEMS ,
)
SURVEILLANCE REOUIREMEtiTS i
4 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
. a. At least once per 31 days on a STAGGERED TEST BASIS by:
1. Verifying that each HPI pump:
a) Starts automatically on a test signal.
b) Develops a discharge pressure of 1 1170 psig on recirculation flow.
c) Operates for at least 15 minutes.
2. Verifying that each LPI _ pump: -
a) Sta'rts automatically on a test ' signal, b)' Develops a discharge pressure of 1 150 psig on
recirculation flow. -
c) Operates for at least 15 minutes. *
$ ~ Verifying that the following va.1ves are open with power to the valve operator disconnected by removal of the breaker from the circuit: ,
Valve Number Valve Function
a. a. -
b. b.
.. c.
. Verifying that the LP injection cross-over valves are
[ 0 {. locked, sealed or otherwise secured in their throttled position.
5. Cycling each testable remotely or automatically power operated valve through at least one complete cycle of full travel.
6. Verifying the correct position for each manual valve not
. locked, sealed or otherwise secured in position, i
l .
j ~ DAVIS-BESSE, UNIT 1 3/4 5-4 July 23, 1975
s i EMERGENCY CORE COOLING SYSTEMS
. ECCS SUBSYSTEMS - T,yq < 280*F l
LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
w; g pc.s presaure > 300 P'ES3
a. g ne OPERABLE-high pressure injection (HPI) pump, j
b. One OPERABLE low pressure injection (LPI) pump, f
c. One OPERABLE decay heat cooler, and
d. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) and transferring suction to the containment amergency sump. ;
l l
APPLICABILITY: MODE 4.
ACTION: .
. a. With no ECCS subsystem OPERABLE because of the inoperability
. of either the HPI pump or the flow path from the borated l
. water storage tank, immediately restore at least one ECCS subsystem to OPERABLE status or be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .
b. With no ECCS subsystem OPERABLE because of the inoperability
~. .
of either the decay het t cooler or LPI pump, immediately
. restore at least one ECCS subsystem to OPERABLE status or maintain'the Reactor Coolant System Tavg less than 280 F by use of alternate heat removal methods.
c. In the event the ECCS is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2
- within'90 days describing the circumstances of the actuation and the total accumulated acteation cycles to date.
SURVEILLANCE REQUIREMENTS j I
4.5.3 .The ECCS subsystems shall be demonstrated OPERA.BLE per the applicable Surveillance Requirements of 4.5.2.
DAVIS-BESSE, UNIT 1 3/4 5-7 July 3, 1975
.l
. . l EMERGENCY CORE COOLING SYSTEMS BORATED WATER STORAGE TANK LIMITING CONDITION FOR OPERATIUN 3.5.4 The borated water storage tank (BUST) shall be OPERABLE with:
a. A minimum contained volume of 360,000 gallons of borated water. .
b. A minimum boron concentration of 1800 ppm,
c. A minimum water temperature of 35*F, and A maximum concentration of ( ) ppm.
A APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the borated water storage tank inoperable, restore the tank to QEERABLE status within one hour or be in COLD 5HUTDOUN within the next
@0) hours.
. 30 e r -
SUR'IEILLANCE REQUIREMENTS 4.5.4 The BWST shall be demonstrated OPERABLE:
a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when ambient air temperature
<35'F by. verifying the water temperaturc.
b. At least once per 7 days by':
1. Verifying the water level in the tank, -'
, 2. Verifying the boron concentration of the water.
DAl"S-BESSE, UNIT 1 - 3/4 5-8 May 19, 1975
.. t
I CONTAff01ENT SYSTEMS
. (
i
. CONTAlf! MENT LEAKAGE l
1 LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: .
a. An overall integrated leakage rate of:
1. <L 0 74$o,urs.50percentbyweightofthecontainmentairper at P , 38 psig, or a
2. < L,, 0.25 percent by weight of the containment air per -
74 Hours at a reduced pressure of P t ,19 psig.
b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Type B and C tests when pressurized to Pa-
c. A combined leakage rate of < L for all penetrations identified in Table (3.6-1)~~as secondary containment bypass leakage paths, when pressurized to Pa- .
( APPLICABILITY: MODES 1, 2, 3 and 4.
, ' ACTION:
With either (a) the measured overall integrated containment leakage rate ex eeding 0.75 La or 0.75 Lt, as applicable, (b) with the measured comoined leakage rate for all' penetrations and valves subject to Type B and C tests exceeding 0.60 La, or (c) with the combined bypass leakage
, rate exceeding M9ta, restore the leakage rate (s) to within the limit (s) prior to) increasing the Reactor. Coolant System temperature above 200'F. L g SURVEILLANCE RE0UIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 - (1972):
a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at either Pa, 38 psig, or at Pt ,19 psig, during each i
DAVIS-BESSE, UNIT 1 , 3/4 6-2 December 30, 1974 I l
t CONTAI!5 TENT SYSTEMS
,7 SURVEILLANCE REQUIREMENTS (Continued) ,
.10 year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.
b. If any periodic Type A test fails to meet either .75 La or
.75 Lt, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive
. Type A tests fail to c.eet either .75 L a or .75 L t , a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either .75 La or .75 Lt at which .
time the above test schedule may be resumed.
c. The accuracy of each Type A test shall be verified by a supplemental test which:
1. Confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A "et data is within 0.25 La or 0.25 Lt . .
- Has a minimum durction o'f 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . following a. stabilization period equivalent to the stabilization period for the Type A test, and
3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25. percent of the total measured leakage rate at Pa 38 psig, or Pt , 19 psig.
e d. Type B and C tests shall be conducted at pa (50 psig) at intervals no greater than 24 montis except for tests involving: -
1. Air locks,
2. Penetrations using continuous leakage monitoring systems, l and
3. Valv'es pressurized with fluid from a s'eal system.
e. TJ)ft combined bypass leakage rate shall be determined to be <
0.15,La by applicable Type B and C tests at least once every g,0!f ('24' months except for penetrations which are not individu
, testable; penetrations not individually testable shall be detcrained to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P a , 38 psig, during cach Type A test.
DAVIS-BESSE, UNIT 1 3/4 6-3 May 2, 1975 t~.
. ..l s
C0tlTAlf! MENT SYSTEMS 4 .
3/4.6.2 DEFRESSURIZATION AfiD C00LIfiG SYSTEMS ,
CONTAltiMEtiT SFRAY SYSTEM LIMITIflG C0f:DITIO!! FOR OPERATION _
i 3.6.2.1 Two separate and independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the BWST on a containment spray actuation signal and transferring suction to the containment emergency sump.
APPLICABILI'TY: MODES 1, 2,'3 and 4. -
ACTION:
With one containment spray system inoperable:
a. Restore the inoperable spray system to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STAf;DBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; restore the inoperable spray system to OPERABLE status within the next
. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD S4UTDOWN within the ner'. 2.0 hours0 days 0 hours 0 weeks 0 months , and .
With 3 containment cooling units OPERABLE, t,ithin (2) hears 1
. ' align containment cooling unit 1-3 to the sane essential bus c(fM' as the inoperable spray pamp unless the spray pump is inoperable as a result of an inopered.e essential bus, then align contain-ment cooling unit 1-3 to the OPERABLE esssntial bus.
e SURVEILLAtlCE REQUIREMENTS l
, 4.6.2.1 Each containment spray system shall be demoastrated OPEPABLE:
a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , when one spray system is inoperable,
- by verifying that containment cooling unit 1-3 is aligned to the specified essential bus.
b. At least once per 31 days on a STAGGERED TEST BASIS by:
L 1. Starting each spray pump from the control room,
2. Verifying, that on recirculation flow, each spray pump
. develops a discharge pressure of 1 56 1 ~ psig at a flow of l'300 gpm, -
DAVIS-BESSE, UNIT 1~ 3/4 6-11 -
May 19, 1975 l
. lt
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. l C0flTAlf!MEflT SYSTEf45 C0flTAlf;MEllT C00LillG SYSTEM i LIf4ITIflG C0flDITI0ft FOR OPERATI0ft Tw a m _ f3 3.6.2.2 @ containment cooling units shall be OPERABLE. ,
APPLICABILITY: i'.0 DES 1, 2 and 3. aMg ACTION:
Gc7 -
W '
M- . .
With on (containment cooling unit 5 inoperable:
a. Restoretbin[operableunittoOPERABLEstatuswithin48 hours or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore %
.the inoperable unit to OPERABLE status within.the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWft within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , and
b. If containment cooling unit 1-3 is OPERABLE, within (two) hours align unit 1-3 to the same essential bus as the inoperable unit unless that unit is inoperable as a result of an inoperable essential bus, then align unit 1-3 to the OPERABLE essential bus.
SURVEILLA!!CE REQUIREMEllTS o 4.6.2.2 Each containment cooling unit shall be demonstrated OPERABLE:
a At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , when unit 1-1 or 1-2 is inoperable, g by verifying that unit 1-3 is aligned to the specified essential bus.
b. At least once per 31 days on a STAGGERED TEST BASIS by:
, 1. Starting each unit from the centrol room, l
l
2. Verifying that each unit operates for at least 15 minutes,
3. Verifying that units 1-1 and 1-2 are aligned to receive electrical power from separate OPERABLE emergency busses, and
, 4. Veri 3ying a cooling water flow rate of > gpm to each unit cooler. . .
DAVIS-BESSE, UNIT 1 3/4 6-13 -
May 19, 1975 I {l
=
IS Reason
,, _ . 3.6.4.1 No comument.
(
4.6.4.1 '
The hydrogen analyzers are calibrated with 0% hydrogen in nitrogen and 2.5% hydrogen in nitrogen. + 0.5% allows for a manufacturing tolerance.
f 3.6.4.2 '
Change "A" to !'The" as there.is only one hydrogen purge system filter unit.
I Add " filter unit" where indicated. The hydrogen purge system at DB-1 consists of only the " hydrogen purge system filter unit." The flow through this unit is supplied by the contain-ment hydrogen dilution system blowers which ar1 covered by Spec 3.6.4.3.
The heater is the only device in the hydrogen purge system filter ,
unit which requires essential power.
" filter unit" - see comment on 3.6.4.2.
" heater" - see comments on 3.6.4.3.
4.6.4.2 Add statement at the end of "b" in order that this testing will not be required if the system is not in operation or if tlat portion of the system has been blocked off.
i 3.6.4.3 The titic has been changed from"" Containment Atmosphere Dilution System" to " Containment Hydrogen Dilution and Contain-ment Recirculation Systems" to reflect the fact that DB-1 has two separate systems. This change is also throughout the spec.
4.6.5.9 "e' requires only the valvos in the CHD to be exercised since there are no valves in the CR system.
t l Bases 3/4.6.5 The bases were rewritten to reflect the systems at DB-1 and to reflect the changes in the specs.
l t I
s a
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- - - -- - - d
C0 ITAI!!MEilT SYSTEf ts 3/4.6.4 COMCUSTISLE GAS CO:ITR0L-HYDR 0 Gell A!!ALYZERS ,
f LIMITit;G CO :DITIO!! FOR OPERATI0ft 3.6.4.1 Two separate and independent containment hydrogen analyzers shall be OPERABLE.
APPLICABILIT : MODES 1 and 2.
ACTI0ft: -
With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in HOT STAfiDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
D SURVEILLAf!CE RECUIREMEi!TS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by:. ,
a. Performing a CHA!illEL CALIBRATI0tl using sample gases containing:
tL yu A & a,'Jww
1. MmrWit2Purif".f&i and D. C t o. s
2. 22mt volume percent hydrogen, balance air.
b. Verifying that each analyzer is aligned to receive electrical power from separate OPERABLE essential busses.
i .
DAVIS-BESSE, UtiIT 1 3/4 6-18 May 19,1975 e 1 a.
C0ilTAll:MEf!T SYSTEMS
( HYOROGEtt PUP.GE SYSTEM LIMITIfiG CO::DITI0ft FOR OPERATIO!!
T*AA.
f &W 3.6.4.2 3 containment hydrogen purge system shall be OPERABLE and capable of being powered from a minimum of one OPERABLE essential bus. A APPLICABILITY: MODES 1 and 2.
ACTIO;1:
)
With the containment hydrogen purge system i3 noperable, restore the hydrogen -
purge system to OPERACLE status within 30 days or be in HOT STA?;DBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . g .
SURVEILLAi'CE REOUIREMEi!TS 4.6.4.2 The hydrogen purge system ~shall be de,monstrated OPERABLE:
a. At least once per 31 days by:
Q
l. Initiatinghlow through the HEPA filter and charcoal adsorber tsese and verifying that the purge system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heatert on, and M
2. LA Verifying that the onrc9yssue is aligned to receive electrical power frem an OPERABLE essential bus,
b. At least once per 18 months or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation and (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal cdsorber l
housings or (3) following painting, fire or chemical release in any ventilation zone communicating with the system,b :g 4
1. Verifying that the charcoal adsorbers remove > 99 percent
' of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI !!510-1975 while operating the purge system at a flow rate of 100 cfm _+ 10 percent..
g m
V -
DAVIS-BESSE, UNIT 1 3/4 6-19 June 4, 1975 e
l I
COITAit:::EllT SYSTE!1S SURVEILLfJ:CE REGUIREF.E!!TS (Contin'u ed)
~
' 2.- Verifying that the HEPA filter banks remove > 99 percent _
of the DOP sehen they are tested in-place in accordance with AflSI !!510-1975 operating the purge system at a flow rate of 100 cfm i 10 percent.
3. Subjecting the carbon contained in at least one test canister or at least two carbon s'amples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifyir.g a removal efficiency of 1 90 percent for radioactive methyl iodide at an air flow velocity of
ft/sec + 20 percent with an inlet mathyl iodide concentra-tion of 0.15 to 0.5 mg/m3 , 1 95 percent relative humidity, and > 190 F; other test conditions shall be in accordance with USAEC RDT Standard i:-16-1T, June 1972. The carbon samples not obtained from test canisters shall be prepared
by either:
(a) ' Emptying or.a entire bed from a removed adsorber tray, '
mixing tha adsorbent thoroughly, and obtaining samples r- -
at least two inches in diameter and with a length ecual v -
to, the thickness of the bed, or (b) Emptying a longitu'dinal sample from an adsorber tray, mixing the-adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. .
4. Verifying a systeit flow rate of 100 cfm i10 percent during system 0;arition.
c. At least once per 18 months by:
1. Verifying that the pressure drop across the combined HEFA
' filters and charcoal adsorber banks is < (6) inches t!ater Gauge while operating the purge system at a flow rate of 100 cfm i 10 percent.
ga
2. Verifying that the air flow distribution to each HEPA filter and charcoal adsorber is within + 20 percent of the l
averaged flow per unit. -
l iar . DAVIS-BESSE, UtlIT 1 3/4 6-20 f
, June 4, 1975 l .
, , ll l
I C0flTAlfMEf!T SYSTEMS AMO cwAsahENT RECIRc0LAT/ cod $'/.$7Eh5
( CO:ITAli:'1ENT AT'.0 SPHERE DILUTIOil M LIMITI!:G C0!:DITIO:S FOR OPERATIOil i
ecahannen+ hydnossa dJuhog 8
3.6.4.3 Two separate and independent e n a ~.. .a wmWW: Mien en l
~~MCid) systems and containment atF recirculation (La4 systems shall be l (cH D) OPERABLE. (cg)
APPLICABILITY: MODES 1 and 2.
ACTION:
cHD cR .,
With one C 3 or c:St system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
SURVEILLAt:CE REQUIREMEllTS CIID CR 4.6.0.3 Each C!S and cut system shall .be demonstrated CPERASLE a- least b once per 31 days on a STAGGERED TEST BASIS by: .
CHD c2
a. Starting ea'ch CAD blower and GA fan from the control room, cHD
b. Verifying that each CAB blower operates for at least 15 minutes and develops a discharge pressure of > 15 psig.
CR
c. Verifying that each CIR fan operates for at least 15 niinutes and develops a differential pressure of > 10" W.G.
can ca
d. Verifying that each EAB blower and GR fan power is aligned to receive eicctrical power frca separate OPERABLE essential busses, and enD
e. Exercising all remotely operated valves in each p flow path through at least one complete cycle of diiG travel. l voAet.
l l
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DAVIS-BESSE, UNIT 1 3/4 6-21 July 3, 1975
. - * * * - = ge
BASES l 3/4.6.5 COMBUSTIELE CAS CO?rrROL r i i
The OPERABILITY of the Hydrogen Analyzers, Containment Hydrogen Dilution System, Containment Recirculation System and Hydrogen Purge System ensures that this equipment will be available to maintain the maximum hydrogen concentration within the containment vessel at or below three volume percent following a LOCA.
i The two redundant Hydrogen Analyzers determine the content of hydrogen within the containment vessel. ,
The Containment Hydrogen Dilution (CHD) System consists of two full cap'acity, redundant, rotary, positive displacement type blowers to supply air to the containment. The CHO System controls the hydrogen concentration by the addition of air to the containment vessel, resulting in a pressurization of the contain-ment and suppression of the hydrogen volume fraction.
The Containment Recirculation System is designed to draw from the areas of -
potentially high hydrogen concentrations in the containment dome and provide a more uniform dispersion of hydrogen throughout the containment vessel. The system consists of two redundant fr,s with independent duct distribution systems.
The Containment Hydrogen Purge System Filter Unit functions as a backup to the ,
CHD System and is designed to release air from the containment atmosphere through a HEPA filter and charcoal filter prior to discharge to the statio'n vent.
S .
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e
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8 @> -9r
C0flTAlf::1EllT SYSTEMS
( SHIELD CUILDIfG IfiTEGRITY t
LIMITIt:G C0!!DITI0il FOR OPERATI0rt 3.6.6.2 SHIELD BUILDIflG INTEGRITY shall be maintained.
APPLICACILITY: MODES 1, 2, 3 and 4. -
ACTI0ft:
Without SHI' ELD BUILDIllG IllTEGRITY, restore SHIELD BUILDIflG IllTEGRITY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in COLD SHUTD0;lil within the next 36 hou s. .
SURVEILLAf!CE REOUIRE"Ef!TS .
, 46 SHIELD BUILDIflG IllTEGRITY shall be demonstrated at least once
%per.6.2 31 days by verifying that:
a.
.c ec6 =/
The door for each.tpenetration room is closed and locked, .
except when being used for nomal transit entry and exit, and
b. Each penetration room blowout panel is closed.
l . a i
I l DAVIS-BESSE, UNIT 1 3/4 6-26 March 10,1975 e
0
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1
. * ** . g 1
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O -
O .1 l
. 1
.+ l 3\
C0f!TAlf:MEllT SYSTE!tS .
V p f.
M..e
. 3/4.6.7 VACUUt1 RELIEF VALVES (OPTIOlAL)
[ . g#
y. -
j ,
LIf41TIf!G CO :DITIOff FOR OPERATI0t! -
i
. 3.6.7.1 The primary containment to atmosphere vacuum relief valves shall be OPERABLE with an actuation set point of ,< _
PSID.
APPLICABILITY: MODES 1, 2, 3 and 4. -
ACTIO;t:
With 'one' primary containment to atmosphere vacuum relief valve inoperable, restore the valve to OPERABLE status.within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to be in COLD SHUTCC'::
within the next 56 hours6.481481e-4 days 0.0156 hours 9.259259e-5 weeks 2.1308e-5 months . -
. , J.,
. . =i SURVEILLAT:CE REQUIREi4ENTS 4.6.7.1 Theprimarycontai[1menttoatmospherevacuumreliefvalves shall be demonstrated OPERABLE: .
a. At least once per 6 months by verifying valve partial opening
(> 5 percent of valve full travel) and that valve operation is not restricted by corrosion, dirt, wear or debris.
b. At least once per 3 yea'rs by verifying that the valves open fully at > PSID. .
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e B&W-DUAL 3/4 6-30L ,
December 30, 1974 i
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( CONTAINME'IT SYSTEMS
- I l BASES 3/4.6.5 COMBUSTIBLE GAS CONTROL 1
The OPERABILITY of the equipment and systems required for the detection l ,
and control of hydrogen gas ensures that this equipment will be avai,lable l to maintain the hydrogen concentration within containment below its flammable 3imit during post-LO:A conditions. Either recombiner unit or the purge system is capable of controlling the expected hydrogen genera-tien associateu with 1) zirconium-water reacti~ons, 2) radiolytic occomposition of water and 3) corrosion of metals within containment.
These hydrogen control' systems are consistent with the recommendations of Regulatory Guide 1.7, ." Control of Combustible Gas Concentrations in .
Containment Following a LOCA." ,
i j
3/4.6.6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSTEP (0PTIONAL 1
I The OPERABILITY of the penetration room exhaust system ensures that radioactive materials leaking from the contair) ment atmosphere through containmen*, penetrations following a LOCA are filtered prior to reaching
(. the environment.- The operation of this system and the resultant affect.
on offsite dosage calculations was assumed in the LOCA analyses.
$Nmuua _ktutr VALVt3 (UF 610ML) ain phere vacuum The OPERABILITY or 6..
relief valves ensures that the co a sure differenti 1 does not become more nej an ) psi. This conditio 3 " "5 ry t'o prevent' c4rrT~dhe containment design limit for internal cressure D FM) p:,i .
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3/4.6.8 SECONDARY CONTAINMENT ,
3/4.6.8.1 VENTILATION SYSTEM .
The OPERABILITY of the shield building emergency ventilation systems ensures that containrent vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the safety analyses and limit the site boundary radiation doses to within the limits of 10 CFR l t 100 during LOCA conditions. ,
t .
May 7. 1975 B&W-DUAL , B 3/4 6-4L ,
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9 C0!frAINMEfff SYSTEMS SHIELD BUILDING INTEGRITY T m TING CONDITION FOR OPERATION 3.6.8.2 SHIELD BUILDING INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
Without SHIELD BUILDING INTEGRITY, restore SHIELD BUILDING INTEGRITY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in COLD SHUIDCWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
1 SURVEILLANCE REQUIREMENTS 4.6.8.2 SHIELD BUILDING INTEGRITY shall be decionstrated at least once per 31 days by verifying that the airtight doors and the blowout panels listed in Table 4.6-xx are closed except when the airtight doors are being used for normal-transit entry and exit.
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O 3/4 6-34L
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l Table 4.6-xx '
- ACCESS OPENINGS REQUIRED TO BE CLOSED TO ThSURE SHIELD BUILDING INTEGRITY y-I. AIR TIGHT DOORS h
_ DOOR NO. DESCRIPTION ELEVATION 100 Access Door from the No.1 ECCS Pump 545' Room (Room 105) to Pipe Tunnel 101 104A Access Door from Stair AB-3 to the 555' No.1 ECCS Pump Room (Room 105) 105 Access Loor from Passage 110A to the 555' area above the Decay Eeat Coolers
107 Access Door from the No. 2 ECCS Pump 555' .
Room (Room 115) to the Miscellaneous Waste Monitor Tank and Pump Room (Room 114) 108 Access Door from the No. 2 ECCS Pump Room 555' (Room 115) to the Detergent Waste Drain Tank at.d Pump Room (Room 125) 201-A Access Door from Corrider 209 to the No.1 565' Mechanical Penetration Roo:2 (Room 208) 704 Access Door from Passage 227 to the 565' Makeup Pump Room (Room 225) 205 Access Door from Passage 227 to the 5 85' No. 2 Mechanical Penetration Room (Room 236) 308 Access Door from Corridor 304 to the 585' No. 4 Mechanical Penetration Room (Room 314)
II. BLOWOUI PANELS TOTAL NO. LOCATION EIEVATION 1 No. 2 Mechanical Penetration Room 565' (Room 236) 6 No. 3 Mechanical Penetration Room 585' (Room 303) 6 No. 4 Mechanical Penetration Room 585' (Room 314) s
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, PLAflT SYSTEMS i
AUXILIARY FEEDWATER SYSTEMS LIMITIfiG C0f'DITIO!! FOR OPERATI0ft 1
3.7.1.2 At least two steam generator auxiliary feedr.:ter subsystems shall be OPERACLE with each subsystem comprised of: ,
a. One OPERABLE auxiliary feedy!atar pump capable of being powered from an OPERABLE emergency bus (OPERABLE steam supply),
b. A separate and independe t OPERABLE flow path capable of taking suction from the condensate storage tank and transferring suction to the -(alternate water source).
APPLICABILITY: MODES 1, 2 and 3. .
ACTION:
i l a. With one auxiliary feedwater subsystem inoperable, restore .
. the inoperable subsystem to OPEP.ABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 'or be in HOT SHUTDOWil within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,
i , .
. SURVEILLANCE REOUIREMEtlTS 4.7.1. 2 Each auxiliary feedwater subsystem shall be demonstrated OPE.RABLE:
a. At least once per 31 days on a STAGGERED TEST BASIS by:
1. Starting each pump from the control room,
-2. Verifyingthateachpumpi
, a) Develops a discharge pressure of > .
psig on l recirculation flow, and b) Operates for at least 15 minutes.
l 3. Verifying that the following valves are open with power i
to the valve operator removed:
Valve flumber Valve Function
a. a.
s. ,
b. b.
c. c.
DAVIS-BESSE, UNIT 1 3/4.7-4 May 2, 1975
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TABLE 4.7-2 SECONDARY COOLANT _ SYSTEM SPECIFIC ACTIVITY SAMPLE At;D AtlALYSIS PROGRAM .
TYPE OF MEASUREMENT .
AND ANALYSIS MINIMUM FREQUENCY
. 1. Gross Activity Determination 3 times every 7 days with a maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples
2. Isotopic Analysis for DOSE bei ?I-@ -
EQUIVALENT I-131 Corcentration a) 1 per 31 days, whenever
. the gross activity determinac tion indicates iodine concenc trations greater chan 105 of the allowable limit.
b) 1 per 6 months, whenever the gross activity determin -
tion indicates iodine con-centrations below 10% of the allowable limit.
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DAVIS-BESSE, UNIT 1 3/4 7-8 July 3, 1975 u.
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PLANT SYSTEMS
(
MAlft STEAM LIf;E ISOLATION VALVES LIMITit;G C0liOITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
With one main steam line isolation valve inoperable, be in HOT SHUTD0!!il .
within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
SURVEILLA!;CE REQUIREMENTS ,
. 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by performance of the following surveillance requirements at alternate 6 week intervals: ,
a. Part-stroke exercising the valve, and -
O
b. Verifying full closure'within seconds on any closure actuation signal at the maximum steam flow permissible
,p without either actuating the secondary system safety valves, initiating a turbine trip or initiating an ESF actuation, or at the existing steam flow if in MODES 2 or 3.
DAVIS-BESSE, UNIT 1 -
3/4 7-9 March 15, 1975 .
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l PLAtlT SYSTEMS
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3/4.7.5 ULTDtATE HEAT SINK (OPTI0tlAL)
LIMITING C0t!DITI0tl FOR OPERATI0ft 3.7.5.1 The ultimate heat sink shall be OPERABLE with:
, a. A minimum water level at elevation 561.6 feet, International Great Lakes Datum, and .
b. A maximum water temperature of < (9 6)fF.
APPLICABILITY: MODES 1,,2, 3 and 4. .-
ACTI0ti:
With'the requirements of the above specification not satisfied, be in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
SURVEILLANCE REOUIREMEtlTS 4.7.5.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the water temperature and water level to be within their limits. '
e DAVIS-BESSE, UNIT 1 . 3/4 7-13 May 19,197S c.
TS Reason
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3.7.6.1 No comment.
l 4.7.6.1 "a" has been moved to the 31 day testing. The Control Room l Emergency Ventilation System is not normally in operation.
During normal operation, the Normal Control Room Ventilation System maintains the control room temperature. (see DB 1 FSAR, Section 9.4.1) . Therefore, the capability to cool the control room will be done every 31 days.
There are no heaters on the filters. "b" should only require testing if that portion was in service.
"C3"" slight" - see DB-1 FSAR, Section 9.4.1.3.
The Control Room Emergency Ventilation System is not automatically started on any signal. The Normal Control Room Ventilation System is automatically isolated by the signals as listed and then the operator must then manually start the Control Room Emergency Ventilation System.
B 3/4.7.7 Revision indicates that the temperature requirement of the system is only under accident conditions.
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PLAtlT SYSTE!'S 3/4.7.6 C0!: TROL R00'1 E!'ERGE!:CY VEr!TILATIOil SYSTE'4 LI!4ITIt!G C0!:DITI0tt FOR OPERATIOff l .
3. 7. 6.1 Two independent control room emergency ventilation systems shall be OPERABLE. ,
APPLICABILITY: MODES 1, 2, 3 and 4. ,
. ACTI0ft: .
With one control room emergency ventilation system inoperable, restore -
the system to OPERABLE status within 7 days or be in COLD SHUTDOW1 within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . '
( -
SURVEILLAt:CE REOUIREME!!TS
~
4.7.6.1 Each control room emergency ventilation System shall be y demonstrated CPERABLE:
cc1K.. - - L - ?fM6 am @ cca mSyn WT ehM-ongp_N1WP.
re . - ~ -
a dr. At least once per 31 days on a STAGGERED TEST BASIS by:
l 1. Initiating flow through the HEPA filter and charcoal l
adsorber train and verifying that the train operates for
at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months h, and I
2. Verifying that each ventilation system is aligned to 1)' Q ALA4 W w,. receive elect 7 cal . power from separate OPERASLE essential 1
A 1 SC 3 4 g,g g & o, At least once per 12 months. or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation and (1) after each complete or partial replacement c b M, of a HEPA filter or charcoal z.dsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber ~
housings,or(3)followingpainting,fireorchemicalreleaye in any ventilation zone communicating with the system b! n 4
.il-L f f A =fW 5 5 -y -l 7
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DAVIS-BESSE, UNIT 1 -
3/4 7-14 March 17, 1975-Ib
PLAtlT SYSTEMS SURVEILLAt:CE REOUIRE!'ENTS (Centinued)
1. Verifying that the charcoal adsorbers remove 1 99 percent
~
of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with AflSI H510-1975 while operating the ventilation system at a flow rate of 3300 cfm i 10 percent.
2. Verifying that the HEPA filter banks remove 1 99 percent of the DOP when they are tested in-place in accordance
, with Af;SI l'510-1975 while operating the ventilation system -
at a flow rate of 3300 cfm i 10 percent.
3. Subjecting the carbon contained in at least one test
' canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal efficiency of 1 90
, percent for radioactive methyl iodide at an air flow velocity of it/sec + 20 percent with an inlet methyl iodide concentration of 0.05 to 0.15 mg/m3 , t 95 percent relative humidity, and > 125'F;'other test conditions ,
f, 'shall be in accordance Eith USAEC RDT Standa'rd M-16-lT, i V June 1972. The carbon samples not obtained from test '
canisters shall be prepared by either:
la) Emptyirg one entire bed from a removed adsorber tray, mixing the aJsorbent thoroughly, and obta.ining i samples at least two inches in diameter and with a length equal to the thickness of the bed, or e
b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining .
. sampics at laast two inches in diameter end with a length equal to the thickness of the bed.
4. Verifying a system flow rate of 3300 cfm i10 percent during system operation.
c. -
6. At least once per 18 months by:
1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < (6) inches Water i Gauge while operating the ventilation system at a flow t
rate of 3300 cfm + 10 percent.
l 1
DAVIS-BESSE, Ut!IT 1 3/4 7-15 June'4, 1975 A A
(' O 1)t ; L*LLwdOUtum &A$ % . - -- - - -
tLkJ. k .t/ N OkbbA<.......----
.. . .. . us .1 % 4. SFAS. ,
. _ . - - - ---. L G.& C - }lg'N9 = 0, p~elo W Y A k 0 . kQ $; Y .-.- . .
. . . - . . . . . . . - - . - - . v. . . . --. ...
I
3. Verifying that the system maintains the control room at a ..._ _..
-- - - - - - mE g li positive pressure C&~w, men m. relative to the outside atmosphere during system operation. . _-- - -._ _...
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.q g a- e. ee e & h g-4 .a. emq m mg, g N
. *666O M65=*6.66 M8-.Au4. .< .N6 _M . W y. -
M g.em p
e. Oem .- e.. 6.MN.6. e. 6.w.m . . MMwb > , e we< (
O* ..45b*6 .hMcae. O e - .e . w _ - e e e.m w mg - em- -
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PLANT SYSTEMT,
-( .
-; BASES The limitations on minimum water level and maximum temperature are 4
based on providing a 30 day cooling water supply to safety related equip-ment without exceeding their design basis temperature and is consistent
- with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants."~
3/4.7.6 FLOOD LEVEL (OPTIONAL) O The limitation on flood level ensures that facility operation will be terminated in the event of flood conditions. The limit of evaluation
( Mean Sea Level is based on the maximum elevation at which facility fEo)d control measures provide protection to safety related equipment.
3/4.7.7 CONTROL ROOM VENTI TION SYSTEM 4
The OPERABILITY of the control room ventilation system ensures th d
(
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1) the ambient air temperature does not exceeo the allowable temperature # . J for continuous dutf rating for the equipment and instrumentation cooled' t#'
' by this system and 2) the control room will remain habitable for opera-tions personnel t ^ g :.J ~f.L....3 ;' :md Mc -h' ~ A#t':S.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 10 of Appendix "A",10 CFR 50.
3/4.7.8 ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM (OPTIONAL)
'/he OPERABILITY of the ECCS pump room exhaust air filtration system ensures that radioactive materials leaking from the ECCS equipment within the pump room following a !.0CA are filtered prior to reaching i the environment. The operation of this system and the resultant effect '
! on offsite dosage calculations wae assumed in the safety analyses.
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B&W-STS , B 3/4 7-4_ ,
. May 7, 1975 e
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F-
. -PLANT SYSTEMS
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3/4.7.7 SEALED SOURCE CONTR11 NATION s
LIMITING CONDITION FOR OPERATION a
. 3.7.7.1 Each sealed source -]ntaining radioactive material in excess of those quantities of byproduct material listed in 10 CFR 30.71 or 0.1 micro-curies, including alpha emitters, shall be free of > 0.005 microcuries of removable contamination.
APPICABILITY: At all times.
~
ACTION: -
a. Each sealed source with removable contamination in' excess of the above limit shall be immediately withdrawn from use and:
1. Either decontaminated and repaired, or -
l 2. Disposed of in accordance with Commission Regulations.
b. The provisions of Specificatiens 3.0.3 and 3.0.4 are not
- applicable.
. SURVEILLANCE REOUIREMENTS -
4.7.7.1.1 Test Requirements - Each sealed source shall be tested for.
. leakage and/or contamination by: -
a. The licensee, or
b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 l microcuries per test sample.
1 4.7.7.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequency descirbed below.
,y
a. Sources in use (r(c,1uding startup sources previously subjected to core flux) - ac least once par six months for all sealed sources containing radioactive material:
DAVIS-BESSE, UNIT 1 3/4 7-17 March 15,1975
ELECTRICAL POUER SYSTE!!S .
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SURVEILLAf:CE REQUIREMEllTS (Continued)
+
, 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
a. At least once per 31 days on a STAGGERED TEST BASIS by:
1. Verifying the fuel level in the day fuel tank,
2. Verifying the fuel level in the fuel storage tank,
3. Verifying that a sample of diesel fuel from the fuel storage tank is within the acceptable limits specified in -
Table 1 of ASTM D975-68 when checked for viscosity, water and sediment,
4. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank,
5. Verifying the diesel starts from ambient condition,
' ~
6. Verifying the generator is synchronized, loaded to > 2000
. kw, and operates for > 60 minut6s, and
7. Verifyiilg the diesel generator is aligned to provide standby power to the associated essential busses.
b. At least once per 18 months during -hutdown by:
1. Subjecting the diesel to an inspection in accorda'nce with
, procedures prepared in conjunction with its marufacturer's recommendations for this . class of standby service, *
2. Verifying the generator capability to re.iect a load of kw without tripping,
3. Simulating a loss of offsite power in conjunction with a
, safety injection actuation signal, and:
! e n,en No /
a) Verifying de-energization of the emer-gency busses and load shedding from the emer-geRy busses, i
e s s e >, # c n / .
b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency-e._.re m l busses with permanently connected loads, energizes the auto-connected emer;;;r.cy- loads through the load sequencer and operates for > 5 minutes while its generator is loaded with the emergency loads.
( cseen*.=/
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DAVIS-BESSE, UNIT 1 3/4 8-3 e_ sse ,. +. ad March 10, 1975
!(
. REFUELING OPERATIO!!5 .," .
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( INSTRUMENTATIOil LIMITING CONDITIO!l FOR OPERATION i .
3.9.2 As a minimum, two source range neutron flux monitors shall be opera.tinbeach_ysith continuous visual indicafionEthe control room %g
, (andonewithaudibleindicationinthecontainment- 4 .
APPLICABILITY: MODE 6. -
ACTIO!!: .
With the requirements of the above specification not satisfied, immediately .
suspend all operations involving CORE ALTERATIONS or positive reactivity changes. .
SUP,VEILLANCE REQUIP.E".ENTS -
4.9.2 Each source range neutron flux monitor shall,be demonstrated OPERABLE by performance of:
a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and
, b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and
c. . A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.
i .
l DAVIS-BESSE, UNIT 1 3/4 9-2 June 30,1975 I -
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REFUELIriG OPERATIOil5 i
FUEL HAtiDLING BRIDGE OPERABILITY ,
LIMITI!!G C0!!DITIOil FOR OPERATION _
3.9.6 The fuel handlin'g bridges shall be used for movement of control {
rods.or fuel assemblies and shall be OPERABLE with:
a. A hoist minimum capacity of 3250 pounds, including weight of grapplel tube, and .
., g .
. b. A' hoist overload cutoff limit of 2806~ pounds, including weight of l grapple tube. .
APPLICABILITY: During movement of control rods or fuel assemb1'es.
ACTION:
With the requirements of the above specification not satisfied, suspend use of the inoperable bridge (s) for all operations involving movement of control rods or fuel assemblies. ,
. SURVEILLANCE REQUIREMENTS 4.9.6 The fuel handling bridges shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of moving control rods or fuel assemblies by performing a hoist load test of at least 3250 pounds and demonstrating an automatic load cutoff when the hoist . load exceeds 28007ounds.
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DAVIS-BESSE, UNIT 1 3/4 9-6 July 23. 1975 i
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MpFcfAL TEST EXCEPTI0flS .
LOW PO'.lER PHYSICS TESTS g ,
LIMITING C0flDITIOlI FOR OPERATI0tl 3.10.3 The limitations of Specifications 3.2.2 and 3.2.3 may be suspended during the performance of PHYSICS TESTS provided: ,
a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and .
. b. The reactor trip setpoints on the OPERABLE Nuclear Overpower channels are set at < DSQof RATED THERMAL POWER.
qs 90
c. The nuclear instrumentation Source Range and Intermediate .
Range high startup rate control rod withdrawal inhibit are OPERABLE.
APPLICABILITY: MODE 2.
ACTION:
With the THERMAL POWER > 54 of RATED THERMAL POWER, imediately manually trip the reactor.
A SURVEILLAtlCE REQUIREMENTS
. 4.10.3.1 The THERMAL POWER shall b'e determined to be < 5% of RATED THERMAL POWER at least once per tour during PHYSICS TESTS by indication on the Intermediate Range and Power Range channels.
4.10.3.2 Each Source and Intermediate Range and Nuclear Overpower channel shall be subjected to a CHANNEL FUtiCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to initiating PHYSICS TESTS.
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DAVIS-BESSE, UNIT 1 3/4 10-4 May 19, 197S c
3/4.1 REACTIVITY CONTROL SYSTEMS j m
BASES 3/4.1.1 BORATION CONTROL ,
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3/4.1.1.1 SHUTDOWN MARGIN ,.
1 A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made
- subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.
During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the minimum
~
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withdrawal limits.
The most restrictive condition occurs at EOL, with Tav operating temperature, and is associated..with-a-postulate,d steam line ga braak accident-and-result ng-uncontrolled i RCS-cooldown-_In. the..ana.lys4s of tttis-accident-a-minimum SHUTDOWN 4tARGIN of-(0.-30)%-ak/k-is~ initially'-
i
. required to control the-reactivity _. transient. Accordingly, the SHUTD0'.lN MARGIN required is-baseLupon_this_1.imiting-condition and is consistent with FSAR safety analysis assumptions.
O 3/4[.l.2 BORON DILUTION .
A minimum flow rate to the core of at least 2800 GPM provides adeqt. ate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2800 GPM will circu-late an equivalent Reactor Coolant System volume of 11,262 cubic feet in approximately 30 minutes. The reactivity change rate associated with i boron concentration reduction will be within the capability fcr operator I
recognition and control. ,
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC) l
! The limitations en MTC are provided to' ensure that the assumptions i
usco in the accident and transient analyses remain valid through eacn l
fuel cycle. The surveillance requirement for measurement of the MTC at the beginning middle and near the end of each fuel cycle is adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
The confirmation that the measured and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides' additional assurance that the coefficient will be maintained within its limits during intervals between measurement.
DAVIS-BESSE, UNIT 1 B 3/4 1-1 July 3, 1975 e
.4- a - -.
l 4 REACTIVITY CONTROL SYSTEMS BASES j- 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY i . This specification. ensures that the reactor will not be made critical #
- with the Reactor Coolant System average temperature less than ( )*F.
This limitation is required to ensure 1) the moderator temperature
. ~ coefficient is within its analyzed temperature range, 2) the pressurizer is capable of being in an OPERABLE status with a steam bubble, 3) the reactor pressure vessel is above its minimum NOTT temperature and 4) the protective instrumentation is within its nonnal operating range.
~
3/4.1.2 BORATION SfSTEMS The boron injection system ensures that negative reactivity control i
is available during each mode of facility operation. The components required to perform this function include 1) barated. water sources, 2) makeup or DHR pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat systems, and 6) an emergency power supply from OPERABLE emergency busses.
l With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection .,ystems are provided to ensure single functional capability in the event an hssumed failure renders one i
of the systems. inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility. safety from injection system failures during the repair period.
~
l The boration capaaility of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 2./ O % ak/k after
. _ xenon decay and cooldown to 200 F. The maximum boration capability l requirement occurs at EOL from full power equilibrium xenon conditions l and requires ther 490 gallons of rm ppm borated water from the boric acid fM s&
a tanks orJo.Eogallons of /Joc' ppm borated water from the borated water storage tank. l The requirements-for a minimum contairled volume of 360,000 gallons of borated water in the borated water storage tank ensures the capa-bility for borating the RCS to the desired level. The specified quantity
- of borated water is consistent with the ECCS requirements of Specification
i. (3.5.4). Therefore, the' larger volume of borated water is specified
!- . here, too. . -
p .
DAVIS-BESSE, UNIT 1 B 3/4 1-2 July 3, 1975 l -
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REACTIVITY CONTROL SYSTEMS t c .
4 BASES l-With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity conditi.on of the reactor and the additional restric-t tions, prohibiting CORE ALTERATIONS and positive reactivity change in' the event.the single injection system becomes inoperable.
' The boron capability required below 200'F is based unon providing a 1% ak/k SHUTDOWN MARGIN at 140*F during refueling with all full and part -
length control rods withdrawn. This condition requires either 76c3 gallons of $7+z ppm borated water from the boric acid stcrage system or 2Lud gallons of geo ppm borated water from the borated water storage ,-
tank.
~
The specifications of this section (1) ensure that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTDOWN i
MARGIN is maintained, and (3) limit the potential effects of a rod ejection i accident. OPERABILITY of the control rod position indicators (LCO 3.1.3.2) is required to determine control rod positions and thereby ensure compliance with the control rod alignment and withdrawal limits.
The ACTION statements which permit limited' variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. For example, misalignment of a safety or regulating rod requires a restriction in THERMAL POWER.
The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exc'eeding the assumptions used in the safety analysis for a rod ejection accident.
- The position of a rod' declared inoper$ble due to misalignment should not be included in computing the average. group position for determining the OPERABILITY of rods with lesser misalignments.
. OPERABILITY of the rod position indicators and control rod positions are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with l more frequent verifications required if anjautomatic monitoring channel r
l e
is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg >_ l (500)'F and with reactor coolant pumps' operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at ' operating conditions.
DAVIS-BESSE, UNIT 1' B 3/4 1-3 July 3, 1975 e
1 r
.l
3/4.2 POWER DISTRIBUTION LIMITS rg g g BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normd Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core
> (1.32/1.30) during nomal operation and in short term transients, '(b) iiiaintaining the peak linear power density <_15.h kw/ft during normal operation, and (c) maintaining the peak powar density < 12M5) kw/ftIn a during short term transients.
met in order to meet the assumptions u' sed for the loss-of-coolant accidents.
The ACTION statements which permit limited variations from the basic .
requirements are accompanied by additional restrictions which ensures that the original criteria are met.
The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:
F Nuclear Heat Flux Hot Channel Factor, is defined as the maximum 0 local fuel rod linear power density divided by the average feel rod linear power density, assuming nominal fuel pellet and rod dimensions.
. F N Nuclear Enthalpy Rise Hot Channel Factor, is defined as the AH ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.
It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on
- minimum DNBR at full power are met, provided:
2 9.+ N Fq_
< (2-r55T; F #
3H - (1.71) l .
Power Peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been determined that the'above hot channel factor limits will bp met provided the following conditions are maintained.
1. Control rods in a single group move together with no individual l
l rod withdrawal differing by more than + M % (indicated position) from the group average height. g l
i DAVIS-BESSE, UNIT 1 B 3/4 2-1 May 7, 1975' e
-.,4 I
POWER DISTRIBUTIO'l LIMITS i i BASES Regulating rod groups are sequenced with overlapping groups as
2. ~
required in Specification 3.1.3.6. l
' 3. The regulating rod withdrawal limits of Specification 3.1.3.6 7
are maintained. .
4. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom halves of the core. Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have .
been correlated with AXIAL POWER IMBALANCE. The correlation shows that the design power shape is not exc,eeded if the AXIAL POWER IMBALANCE is maintained between +_(10) percent'and
- p 5) percent at RATED THERMAL POWER. W
'The desfgn limit power peaking factors are the most restrictive calcu-lated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal and are the core DNBR design basis. Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met. When using incore detectors to make power distribution maps to determine FqandFyH Themeasure' men?oftotalpeakingfactor,FMeas shall be
a. ,
increasedbyf37percenttoaccountformahufacturing tolerancesandfurtherincreasedby(f)'percenttoaccount for measurement error.. y .
N
b. sha
- The measurement of enthalpy rise Hot channel factor, Fbe increa S 2d. A -
For Condition II events, the core is protected from exceedin kw/ft locally, and from going below a' minimum DNBR of (1.32/J,.-30)g , by automatic protection on power, AXIAL POWER IMBALANCE, pressure and temperature. Only conditions 1 through 3, above, are mandatory since the AXIAL POWER IMBALANCE is an explicit inhut to the Reactor Protection l System.
1
! The QUADRANT POWER TILT limit as ures that the radial power distribution satisfies the design values used in the power capability l'
analysis. Radial power distribution measurements are made during startup testing and periodically during power. operation.
DAVIS-BESSE, UNIT 1 B 3/4 2-2 July 23,1975 e
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POWER DISTRIBUTION LIMITS .
' PROOF & RdVIEW CO'Y BASES l.04-The QUADRA;1T POWER TILT limit of (,LB57 at which corrective action
- is required provides DNB and linear heat generation rate protection with x-y plane power tilts. A-hmhinetiltaf-(-1.055)- can be tolerated before-the-margin-for--uncertainty -in-F7 s-depleted--
i The-limit of (1.00) was-se-lected- i.v Av% 29 allowance -for the-uncertainty associated-with thA ladicated-power-tilt.- In the event the tilt is not corrected, the margin fe-unceri. inty on Fn is reinstated by reducing the power by 2 percent for each percent of tYlt in excess ofMr
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DAVIS-BESSE, UNIT 1 B 3/4 2-3 May 7, 1975 e
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3/4.3 INSTRUMENTATION
+
BASES
)
STANOBY 3/4 .3.5 REMOTE SHUTDOWN INSTRUMENTATION l
The OPERABILITY of the remote shutdown instrumentation ensures that <
su icient capability is available to permit shutdown and maintenance of l HOT SWTDOMtof the fccility from locations outside of the control room. '
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.
c.oto man cally SA s dowa anc/
3/4.3.3.6 CHLORINE DETECTION SYSTEM l A -
norm.I -
The OPERABILITY of the chlorine det, etion systems ensures that an accidental chlcrine release will be detected oromittJ1and the ontrol room mcacy ventilation system willeTtc=Ltf ~'bHsolatedthe centr ^1 m.- .nu . .. i t : a tu lL vri u: 71n the recirculation mode to provide
~
the required protection. Thc T hlorine detection ry +am ran" ired by
+hi :peci#icctier crc cen;f ctEnt '>4+h +he rarnmondatinne nf Reg h ry Guide 1.95, "Prctcetien cf 'Mc!bcr Po,ier Plani Cuni.rui Ruvai Operctions
^= inct cn Accidenial Chiviinc Rilcase."
D L ogtu~.3.40t Can t b C Y's
}n :hEle fAC Ina n u a. lly C o ev'* ro l 'b * * * * 'O'^V ye a t-1 I I ct.th on Sys f C m -
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l DAVIS-BESSE, UNIT.1 B 3/4 3-3 July 3, 1975
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pg 4 gg:t dx A I ', 'A EA BASES FIGURE 3-1
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Fiay 7.1975
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B 3/4 3-4 .
B&W-STS I3 ' -
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R ADI AL SYt::ETRY y -
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RADIAL SYMMETRY f
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INCORE lilSTRUMEf1 TAT!0:1 SPECIFICATI0il
. MI!!! MUM RADIAL TILT ARRANGEMENT BASES FIGURE 3-2 I
B&W-STS .
B 3/4 3-5 May 7, 1975
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~ REACTOR COOLANT SYSTEM -
t . .
( BASES
(
1 J2 0 Reducing Tavg to s (500)*F prevents the release of activity should ;
1 a steam generator tube rupture since the saturation pressure of the j primary coolant is below the lift pressure of the atmospherio-steamVQNC3 reLisf-v&ves. (> w .: e + m a.n ha.m i. u. c ode ca}hiy a
i The surveillance requirements provide adequate assurance that l excessive specific actizity levels in the prirnry coolant will be l detected in sufficient time to take corrective action.
-Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency a l
of isotopic analyses following power changes may be permissible if-justified by the data obtained. ,
3/4.4'.9 PRESSURE / TEMPERATURE LIM'TS All components in the' Reactor Coolant S'ystem are designed to with-stand the effects of cyclic loads due to system temperature an.d pressure changes. These cyclic loads are introduced by normal load transients,-
reactor trips, and startup and shutdown operations. The various categories of load cycles used .^or design purposes are pr6vided in Section'( )
of the FSAR. During heatup and cooldown, the rates of temperature and pressure changes are limited so that the maximum specifi.ed heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel ivall proiuce thermal stresses which vary from compressive at the inner wall to vensile at the outer wall. These thermal induced compressive stresses tent to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady state conditions l
(i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of tae vessel is treated as the governing location.
l The heatup analysis also covers the determination of pressu're-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients estab-i lished during heatup produce tensile stresses at the outer wall of the
,' DAVIS-BESSE, UNIT 1 B 3/4 4-5 July 3, 1975
~i
BASES
,, 3/4.6.8.2 SHIELD BUILDING Im'EGRITY t
SHIELD BUILDING INTEGRITY ensures that the release of radioactive material from the 1 containment vessel vfil be restricted to those leakage paths and associated leak rates assumed in the safety analysis. The closure of the airtight doors and blowout panels listed in Tab 1r, 4.6-xx ensures that the Emergency Ventilation System (EVS) can provide a negative pressure between 0.25 to 1.5 inches water gage within the annulus between the shield building and containment vessel and within the interconnecting mechanical penetration rooms after a loss-of-coolant accident. This restriction, in conjunction with the operation of the EVS, will limit the site boundary radiation doses to within the limits of 10CFR100 during accident conditions.
8
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At the present time, the DB-1 annunciator and computer descriptions for the RPS trips do not agree with the nomenclature used in the B & W STS. Various procedures b,,i and training information have been written using the trip names as found on the annunciator and in the computer. If DB-1 was. to be consistant with the B&W STS nomenclature, it would require backfitting since the annunciator windows would have l to be re-made, the computer software would have to be updated, and the procedures and training information would have to be revised. This would be a great expense i
i to TECo since it would require backfitting. Therefore, TECo proposes changing the B&W STS nomenclature as shown below. This nomenclature should be changed in all applicable specs and their bases.
B&W STS CHANGE TO Nuclear Overpower -
High Flux RCS Outlet Temperature - High RC High Temperature Nuclear Overpower Based on RCS Flux-DFlux - Flow Flow and AXIAL POWER IMBALANCE RCS Pressure- Low RC Low Pressure .
RCS Pressure - High RC High Pressure RCS Pressure - Variable Low RC Pressure - Temperature Nuclear Overpower Based on Pump High Flux / Number of Reactor Coolant Monitors Pumps On Reactor Containment Vessel Pressure - Containment High Pressure High Shutdown Bypass RCS Pressure - High Shutdown Bypass High Pressure
,- Action 2 & 3: Only one channel may be bypassed at one time due to a RPS internal electronic interlock. The only way to do the testing is to reset the
, inoperable channel.
Consnents on Sections 8, 9, and 10 have been forwarded showing the needed changes.
e 9
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TABLE _2.2 .
TRIP SETPOINTS_
REACTOR PROTECTION SYSTEM INSTFUMENT'ATIO ALLOWABLE VALUES TRIP SETPOINT
. FUNCTIONAL UNIT Not Applicable
' not Applicable I 1. Manuc1 Reactor Trip <( )% of RATED THERMAL POWER 4 A'- FM Nu6(ear-Overpower (1) i 105.5% of RATED THERMAL POWER s 2. 1 RC H u' Tm A <( )*F ,
'i
3. RCS-Ou et-Temperature higha < 619 F '
FM ~ DFk ~ Fhaa*C& 1.08 x flow with a reduction (3) ,
4. Hurlear-.0verpowerJ ased-on due to AXIAL POWER IMBALANCE .
RCS Elow_and. AXIAL--POWER as shown in Fig w e 2.2-1 UGALAllCE-(@)-
t .
lL . ?~}. . ;
N When the muc1') car Overpows trip setpoint is required to be reduced by an ACT ro bination, '
O' to some percentage of the THERMAL POWER allowable for the reactor coolant pump
. the THERMAL POWEB allowable: '
a.
b.
For3.pumpoperation,is(78%)ofRATEDTHERMALPOWER.For ope t:T 5
(2) Trip may be manually bypassed wien RCS pressure < 1820 t
O psi IW 5% of RATED THERMAL POWER.
trip set point is < 1820 psig andO
a. The Nuclear Overpove rip set point is 1 Pressure L111gh The Shutdown Bypass y removed when RCS pressure > 1900. 4 b.
c. The Shutdown Bypass is b
-)
(3)The channels' ALLOWABLE VALUE trip set point shall not exceed.rq its c
",. than (4) percent.-
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TABLE 2.2-1.(Continued) 5 REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS s -
TRIP SETPOINT ALLOWABLE VALUES M
FUNCTIONAL UNIT
- Rc. Leus % g 1800 psig 1( )psig c: 5. RGS-Pressure-1:ow(2) 5
~'
6.
Rc Hieh W RGS-Pressura-High < 2355 psig <(
~
)psig R c, fks % ~ L ; w o h 4.
g 16.25 Tout F - 7873 psig g((16.25) tout F())psig
7. RCSPxessure-Varicble_1.os(2) fl u h F Q / w s m of. %55% of RATED THERMAL POWER < ( )% of RATED THERMAL POWER
8. Nudlear-Overpower-Based With one pump operating in Honitors- With one pump operating in ey CO
- on-pumpw M -
9 -6 (4 P(23 each loop.
each loop. ,
(0.0)% of RATED THERMAL J < 0.0% of RATED THERMAL POWER With no pump operating in either F0WER with no pump operat-
. loop. ing in either loop.
Ce daw w - L h w H Q *
< 4 psig < ( ) psig
9. Reactor-Containment-Vessel- _
W o .
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. PROOF & REVEV/ CO?(
c 8 i w 2.2 1.IMITIllG SAFETY SYSTEM SETTItiGS BASES 2.2.1 REACTOR PROTECTI0tl SYSTEM It!STRUMEttTATIOff TRIP SETPOI
The Reactor Protection System Instrumentation Trip Setpoint specified
~
in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are pr2 vented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its speci ~ied Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip used in safety analyses.
l Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic ,
Reactor Protection System instrumentation channels and provides manual reactor trip capability.
O a m r4;;w.mr
~
d4 FM A kcica Ciccmcr trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from rapid reactivity excu.~sions .
- During normal station operation with all reactor coolant pumps aperating, reactor trip is initiated when the reactor power level reaches 105.5% of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be (112)"., which was Jsed in the safety analysis. -
R C. H & T y %
2CS Out M t Tm...perai.m c - ::;gh ,
The 0 0t t ::ig trip 1. 619*F prevents the reactor autlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.
O ,
April 25,1975 DAVIS-BESSE, UNIT 1 B 2-4
- l PROOF & RGV;EW COPY LIMITING SAFETY SYTEM SETTIrlGS BASES l
FM - D Fk - Pkw-Lcic:r Omr=.a Ce J sr. "CC "i:- ..J A" AL PC'.CR Ls i A R The power level trip setpoint produced by the reactor coolant system
. flow is based on a flux-to-flow ratio which has been established to accomodate flow acreasing transients from high power where protection i is not provided "r +% clear O x rpcwc M r ed e S
'tcrs ch.... @ .
H 5%/% ,f P =%
The power .- 'ir setpoint produced by the flux-to-flowDn ratio 9 3rovides both h., -
level and low flow protectio.1 in the event the -
reactor power level n.oreases or the reactor coolant flow rate decreases.
~
l The power level setpoint produced ty the flux-to-flow ratio provides averpower DNB protection for all modes of pump operation., For every flow ate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follcws:
l
1. Trip would occur when four reactor coolant pumps are operating
-~
if p:/. r if 100.0% and reactor ficw rett is 100%, or .'ict., ~
) rate is 92.6% and power level is 100%.
2. Trip would occur when three reactor coolant pumps are operating if power if 80.7% and reactor flow rate is 74.7%, or flow rate is 69.4% and power is 75%.
3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.9%
and reactor flow rate is 49.0% or flow rate is 45.4% and the power level is 49.0%.
l For safety cal ulations the maximum calibration and instrumentation errors for the power level were used.
The AXIAL POWER IMBALANCE boundaries are established in drder to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The AXIAL POWER
MBALANCE reduces the power level trip produced by the flux-to-flow ratio l c uch that the boundaries of BASES Figure 2.2 are produced. The flux-to-l
' low ratio reduces the pcwer level trip and associated reactor power-r eactor power-imbalance boundaries by 1.08% for a 1% flow reduction.
l
b. -
(AVIS-BESSE , UNIT .1 B2-5 April 23,.1975
/
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PROOF & REVEW' CO Y -
[ L 1.IMITING SAFETY SYSTEM SETTINGS O
BASES Pm - W ' ~=% '
l
- RC$ Pressure - tow, High and ":
M e ' -
During a slow reactivity insertion s urtup accident from low m ighy P e fr power or a slow reactivity insertion,aegp y Ovefmmr high power, the trip setpoint. The RC$ oreur setpoint is reached before the Hett-Td::.g
?c;Lm m 2355 psig has been established Mtrip any setting limit fM4:.'a~Tuaintain tne system pressure below the safety design transient. The PCS Pcm m . ..;gh trip also backs up the -
/
.Ac H y P W a
RC. Hg nuc 22- e:rpe.:r trip. .
ac. P p_ %k The w Ie:- neT. mun m m-Lew w 1900 psig and re: I. - ; m -Gimr:=23c Lee 16.25
- Tout *F- 7873 psig trip setting limits have beenfor established those design to maintain accidentsthe DNB ratio greater than or equal to (1.32/1.30)It also prevents reactor operatio that result in a pressure reduction.
i at pressures below the valid range of DNB correlation limits, protecting against DNB.
Due to the calibration and instrumentation-errors, the safetyTout -
-:J..SC m
-wrn analysis used a E**cc- upt u+w
4-l~
a. t. rip setpoint. of 15.2
F- (7923 psig).
IIG FM/%4
%~~ %_ ou _L
~ -
- L .6At.
o._.. Pys om k.ppu . n &FkIL!Q v.a r..m In conjunction with the pc;;cr, -timn.e/ .-'rew trips the U c.,
yW Sased-On Mp-?bnitas4 trip prevents the minimum core DNBR from decreasing below (1-.32/1.30) by tripping the reactor due to the loss of reactor pT g coolant pump (s). The pump monitors also. restrict the power level for
~
~
the number of pumps in operation. ,
MN.u$4 W Re ctcr ChieisentWessel " Zur:
l l
N #dA Pe%
The Rconc. Cungthmsa '/cz:! " :=m - .Mgh trip setting limit
< 4 psig provides positive assurance that a reactor trip will occur 'n ]
I the unlikely event of a steam line failure in the containment
'nnm- vessel - L or v.c a loss-of-coolant accident, even in the absence of a RC:
trip.
gc Q pp June 24, 1975 f
DAVIS-BE'iSE, UNIT 1 B 2-6 l
l
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. PROOF & REV!EW CO?Y
(
9 0
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l l
FQ - DF& - f% T*
AYI". P0"[F I"[A'.ANCE rwuvu;ea te Dmar/F10W Ratin Tr% '
BASES Figure 2.2 l
i DAVIS-BESSE, UtlIT 1 B 2-8 Ap 1 23, 1975 .
e .
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POUER DISTRIBt"I0t! LIMITS PROOF & REV!EW COiY HOT CHAffilEL FACTOR - Fq LIMITIrlG C0tlDITI0tl FOR OPERATI0ft j 3.2.2 F shall be limited by.the following relationships: i q ,
I p )- for P > 0.5 . ;
F q1 .
1 2( ) for P 1 0 5 THERMAL POWER' 0.
and P 1 1 where P = RATED THERMAL POWER APPLICABILITY: MODES 1 and 2*.
. ACTI0ft: '
With Fq exceeding its limit:
a.
within its (imit, or Immediately bring F.0
g N !
b. Immediately reduce THERMAL POWER and the "x .a r 5 :r p.;r trip setpoint in direct proportion to the excess that F9 l exceeds its limit, and
c. Demonstrate through in-core mapping that F is within limits :
within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY withik the next ,. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . l i
i SURVEILLANCE REQUIREMENTS 4.2.2 F shall be determined to be within its limit by using the incore g -
! detectors to obtain a power distribution map:
a. . Prior to initial operation above' 75 percent of RATED THERMAL -
POWER after each fuel loading, and.
l
b. At least once per 31 Effective Full Power Days.
See Special Test Exception 3.10.4.
} DAVIS-8 ESSE, UNIT 1 3/4 2-3 . June 4, 1975 3
k
PROOF & Ra/EN COPY ;
II POWER DISTRIBUTIOff LIMITS. *
' H OTCHAtlNELFACTOR-Fk .
l L
IMITIrlG CONDITI0ft FOR OPERATION l hip:
shall be limited by the following relations 3.2.3 F H f
4H 1( l[1+0.2(1-P)] .
THERMAL POWER T
where P = RATED THERMAL P0h R and P < 1.0 APPLICABILITY: MODES 1 and 2*.
ACTION: ,
exceeding its limit: tion to the l With g (Q a. Immediately reduce THERMAL POMR in direct pr excessthatFyg 4 trip setpoint.
reduce the p qNuclear
/t w 4ver W 9 3g is within limits b.
Demonstrate through in-core mapping that F t 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY w a.
I 'RVEILLANCE REQUI_REMENTS ing the shall be determined to be w' thin : its' limit by us 4.2.3 F H '
incore detectors to obtain a power distribution THERMAL POWER map a.
Prior to operation above 75 percent of RATED c after each fuel loading, and l
t At least once per 31 Effective Full Power Days.
i b.
See Special Test Exception 3.10.4.
June 4, 1975 3/4 2 A ,
DAVIS-BESSE, UNIT 1
,j 4
,1
POWER DISTRIBUTION LIMITS PROOF & P.i.VEW 'CO?Y l O QUADRANT POWER TILT _
' - LIMITIllG COMOITIOlt FOR OPERATION 4
3.2.4 The QUADRANT POWER TILT shall not exceed (5)%.
APPLICABILITY: MODE 1.*
ACTION:
With the indi~cated QUADRANT POWER TILT determined to a.
)%,
(5)% but < ( _
1. Immediately correct the power tilt, or
2. Reduce THERMAL POWER so as not to exceed THERMAL POWER, including power level cutoff, allowable for the reactor coolunt pump combin: tion less two percent for every percent of indicated QUADRANT g' pk POWER TILT, andn A Pk -O% FAsw Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the H " eaw 4' erecuer - h iuc4e w
\ 3. - G eE u naAt+.sEe O -
c.-. s ..-. = :ce er " z r w . = e ^.x: .
trip setpoint 2% for every percent of indicated QUADRANT
. POWER TILT.
b. With the indicated QUADRANT POWER TILT determined t
( -)% or exceeding (5)5 for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , immediately reduce THERMAL POWER to (50)% of THERMAL POWER the reactor coolant pump combination and within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months g 7Euce the l?ucler4/crpow+c trip setpoint to < (55)% of _
THERMAL POWER allowable for the reactor coolani, purnp combination be in HOT STANDCY within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SUR"EILLAMCE REQUIREMENTS _
l 4.2.4 The QUADRANT POWER TILT shall be' determined to I
a. At least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during operation above 15% of RATED THERMAL POWE'l except when the QUADRANT POWER TILT is inoperable, thg at least once per hour. *
. e
See Special Test Exception 3.10.4.
O DAVIS-BESSE, UNIT 1
+
3/4 2-5 - April 23, 1975 s
e
-s F O -
o .
- O -
TABLE 3.3-l_ .L REACTOR PROTECTION SYSTEM INSTRUMENTATION _
g MINIMUh !.
CHANNELS APPLICABLE .
(,
G TOTAL N0. CHANNELS OPERABLE- MODES _ ACTION _
a TO TRIP -
OF CHANNELS-G FUNCTIONAL U1IT -
2 1, 2 and **
1 ?
2 1 i
1. Manual Reactor Trip '
3 1, 2 2 '
c= H r- K F b .,e 4 2 )
5 2. er 9 3 1, 2 3 *
, t Nuclean 1'OvepTe-- High- 4 2 t,
3.- kbb([uT1dtT'MEkb r,h e, pi2.s -
2
4. JJuclear-Overpower-Basud-m 3 1, 2 4
2(a) '
Flow e nd-AXIAL-P.WER.. 0 IMBALANCE .
1, 2 3 2(a) 3 R c. Lou Pou:.m<- 4 3
5. RGS-Pressure-L-ow 2 3 1, 2 4
6. kb- re sh5tiigh 1, 2 3
~
2(a) 3 !
' ca 7.
~
^'
"kfkN '4 . -
% ha sh 0abl$ }& tf & .r? ku$ k. W 1, 2 3 l
8. Hublear-4verpowerdase.d on 4 . 2(a) 3 3, r Rug 4tonitor Po.p os 3 - 1, 2
[ 'ifgit .4 , 2 '
9. . hgogt neg g sscc 7; l
4
10. Intermediate Range, Neutron Flux 2 0 ,
2 1, 2 and
O
- and Rate O Source Range, Neutron Flux and Rate .
2# and
5 -n
11.
0 2 2
A. Startup 3, 4 and 5 6 C,.o 0 1 2
B. Shutdown 3 1, 2 and
7 .u
, 4 2 '
12. Reactor Trip Module Logic 3, 4 and 5** 6 ..y 1-2 3
13. Shutdown Bypass RCS:Pressucc ::!yi 4
'H PW G
g n
O a
4
__ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ k
. PROOF & REVEW COPY
( TABLE 3.3-1(Contnued)
J
TABLE NOTATION _
With the Reactor Protection System trip breakers in the closed position !
and the control rod drive system capable of rod withdrawal. i
, j
When Shutdown Bypass is actuated.
10 amps on both i High voltage to detector may be de-energized above 10 .
l intermediate Range channels: '
1820 psig by (a) Trip may be manually bypassed when RCS pressure 1 actuating Shutdown Bypass provided that:
H4 Pk(1) The 'Mkr 0=rre:~ trip setpoint is 1 5% of RATED THERMAL POWER, M y % "igh trip setpoint is 1 1820 (2) The Shutdown Bypass K " :::ur:
psig, and The Shutdown Bypass is manually removed when RCS ' pressure >
(3) .
psig.
I ACTION STATEMENTS _ .
With the number of channels OPERABLE.less than required ACTION 1 by the Minimum Channels OPERABLE requirement, restore th inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or open the control rod drive trip breakers.
ACTION 2
- With the number of OPERABLE channels one less tha Total Number of Channels and with the THERMAL l a.
< 5% of RATED THERMAL POWER, immediately place the-Inoperable channel in the tripped condition; restore the inoperable channel to OPERABLE status prior to l increasing THERMAL, POWER above 5% of RATED THERMAL POWER.
l ,
5 5% of RATED THERMAL POWER, operation may continue b.
provided all of the following conditions are satisfied:
!- 1. The inoperable channel is immediately placed in the tripped condition.
2. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be
bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance Q
v
, testing per Specification 4.3.1.1, cn June 24, 1975
- 3/4 3-3 DAVIS-BESSE, UNIT 1
\
. l PROOF & REVIEW COPY I
' TABLE 3.3-1 (:un unu e TABLE NOTATION and the inoperable channel above may be by -
,fNmed for up to minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when necessary to test the trip breaker l associated with the logic of the channel being tested per Specification 4.3.1.1, and FM RATED
- 3. THERfML POWER is restricted to'< (75)% trip is THEPJML POWER and the """-
reduced to < (85)5 of RATED THERMAL POWER or the QUADRANT P0iiER TILT is monitored at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . J ACTION 3 - With the number of OPERABLE channels one less than the -
Total Number of Channels and with the THERfML POWER level: '
a. < 5% of RATED THERIML POWER, immediately place l l
the inoperable channel in the tripped condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERfML POWER. '
> 5% of RATED THERMAL POWER, op,eration may continue l
b. '
provided both of the following conditions are h '
satisfied: l
1. The inoperable channel is immediately placed.in the tripped condition.
2. The Minimum Channels OPERABLE requirement is met; ,
however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per and the inoperable
' Specification 4.3.1.1,&p^as' channel above may be 4 sed for up to (J5'f 30 minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when necessary to test the trip breaker associated with the logic of the channel being tested per Specification 4.3.1.1. ,-
With the number 6f channels OPERABLE one less than required Action 4 -
by the Minimum Channels OPERABLE requirement and with the THERIML Power level: 1 1
< 5% of RATED THERMAL POWER, restore the inoper- l a.
able channel to OPERABLE status prior to 'ncreasing
' THERMAL POWER above 55 of RATED THERMAL POWER. )
June 24, 1975 i DAVIS-GESSE, UNIT 1 3/4 3-4
- 1 l
3 TABLE 3.3-2 i
g<
~ REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES ! ^
T cs -
RE3PONSE TIMES 0 FUNCTIOilAL UNIT _ !
< seconds
1. Manual React;r Trip E HA Yh4 Htrekean-Overpower *
~
seconds .
q 2. ~j a
RC. H E~l, 'D vfAa i seconds j
3. RES.-Outlet-Temperatupistr !
l FtL4 - D FL4 - Fh ~
tlue-lece-OverpowerBased-on-RGG . ivw auG' l
4. . 1 seconds !
. AXIAL 20MEILIL1MLANCL ,!
Rc. % Pnw i_ seconds !
w 5. ES-Pressure -tm -
R< CA Pm seconds
- 3 RGS-PresureWiigh.
~ 1_
, w 6. ;
Rt Pasaw e =Y u < seconds 6 g
7. Variable-Low-RCS-Prdssure-
.44 Fh / 7tu~h .g M G.l.4 Py os 0 -
l
~"
seconds. 70 '
8. nub 1 ear-Overpower-Based,n Fun,y ",vnii.cn*
- u x k t, ii. M P N m '
i_ seconds -() ;
- 9. Rea c tor-Gentaim66n t-Pres s u re-High -
s ___. nn,4 ()
so. %L2 L S y s Ih } P m Response time shall be T
{p
* !!eutron detectors are exempt from response ti,me testing. measured - ~~0 from gr E'
g Q
Fii
= ,
-C' in" U1 n
-a
o.-
0 O .
TABLE 4.3-1 ,
o REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E CHANNEL MODES IN WHICH
@ CHANNEL FUNCTIONAL SURVEILLANCE g CHANNEL CitECK CALIBRATION TEST REQUIRED g FUNCTIONAL UNIT m N.A.
~
H.A. N.A. S/U(1)
- 1. Ma al FfeV, Beactor Trip
= # D(2), and Q M 1, 2*
S p 2. Nu ,//- ea 0 TPoweL 1 > %M L-RC S R M 1, 2
3. RCS-Out t m I: 'a - D?.h perature--High-FLv -
4. Nuclear 0verpower_ Based-on-RGE M .1, 2*
T-low-and AXIAL _f0WER_lBBN_AttrA S(4) M(3)andQ '
M 1, 2
.5. R eg u $$ bib S R ,
M -1, 2 S R 6.
7.
fts.
Rc Pres'sur$lli}I[&.
Vara ablV m
~) .e s
n LoteRCS-Pressupe S R M 1, 2
o
!;:> 8. Huolear-tL:',
ve/pu.. s s .p%&gW .
rpower-Based 1,2 S R H
on-Pum , ,
w . Har % Ga?p-Honj
.s.~1. tovdh Pimw"igh S R M 1, 2 4 9.. Weactoe-Containmen@Pressurc-- T
10. Intermediate Range, Neutron N.A. - S/U(1)(5) 1, 2 and**
D Flux and Rate S O
11. Source Range, Neutron Flux S N.A. S/U(1)('5) 2, 3, 4 and sO and Rate N.A. N. A. - M(6) ands /U(1) 1, 2 and**

12. , Reactor Trip Module Logic M 3, 4 and 5** *

13. Shutdown Bypass RC-9 /4-i km S - R ..
d
Q,,
Pressure High ,. 2 - 4 '
' Fii.
5 -
~g
., y O 0 . 9c.
3/4.7 PLAtlT SYSTEMS . 3/4.7.1 TL'RDIt:C CYCLE SAFETY VALVES _ LIMITkf!GC0t:DITIO:1FOROPERATI0tl 3.7.1.1 All main steam line code safety valves shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3.
- ACTIO:1:
With one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the "";'02- 0"cr,mc'c^- # - trip :,etpoint is reduced por Table 3.7-1; ot!}erwise, be in COLD SHUTD0;. NF' within the next 36 hours. l .
^
( SURVEILLAtlCE RE0UIREMEtiTS J _. 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice si:cs as shown in Table (4.7-1), in accordance with Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code, dated July 1, 1974. (. - DAVIS-DESSE, UNIT 1 3/4 7-1 February 28,1975
. , I - - . --- . J - . . . - - .
% 4A W
4++/ _ EEm eA1,em TEST TARGET (MT-3)
%4 4 1.0 lf a EM S I ~ m !E==
l,l hdbzb I.8 1.25 1.4 1.6
< 6~ >
MICROCOPY RESOLUTION TEST CHART
#4 4$ '4 '*S$a48
s#/
f 7/~8 o
*s i +
A +e+V
V
& 0 & $g \g/o%'$+>$9#y /v% g /////
pg,4g,'</; ete
%~#'e9, s ,M eE Ev <e 1,em %'+$
TEST TARGET (MT-3) I l.0 !!m EM I y @ EE I.l l '" lillM l.8 1.25 1.4 1.6
} _.
6" = MICROCOPY RESOLUTION TEST CHART
+4 #p 44<>A+e 47% # ,%'7///
+r*d<o +4 ya- j ' l \
+444'
o ~
' ~
g . o o O, TABLE 3.7-1 a A f( Fk - MAXIMUFi ALLOWAC'.E NUCLEAR-0 VERI @lER TRIP SETPOINT WITH INOPERABLE , STEAM LINE SM ETY VALVES 3 Maximum Allowable Nuclear
;p' Mavimum Number of Inoperable Safety Overpower Trip Setpoint - Valves on Any Steam Generator (PercentofRATEDTHERMALPOWER)
E - M ,
._, 1 ( ).
2 ( ) 3 - ( ) ; a - j: b ' n 8 9
? -
ET lif s
- - - -_________L_-
F
]
4. . . .: . . - .
,c - '~
1 ,. . ,,
.Q 1 '3/4.4 REACTOR C00LAflT SYSTEM .
l )j
~
BASES . [fr g
; i -
I 3/4.4.1 REACTOR C00LAtlT LOOPS 9 The plant is designed to operate.with both reactor coolant loops inQ N
'! operatic.1, and maintainWith DNBR abovecoolant (1.32/J.30) during all normal one reactor pump not in oper- bg,o I and anticipated transients.
ation in one or both loops, THERMAL POWER is restricted and th9 M by"-lthe earNedreer j NN @VJ Oveek.w. Zwedwm-RGS-+h =d l'CT.'_ m'n W Af AMCC gge g,7 p ge s en o,,-.? '* Ns trip, ensuring that the DNSR will be < at the maximum possible THERMAL. POWER for } dj l maintained above (1.32/1.30) the number of reactor ccolant pumps in operation or the loL1 quality vhf at the point of minimum DN8R equal to (22/15)". whichever is more ; i . res'trictiVe. - i 8 A single reactor coolant loop provides sufficient heat removal I capability for removing core decay heat while in Mode 3; however,
' single failure considerations require placing a DHR loop into o9 ope in the :;hutdown cooling made if' component repairs and/or corrective i
__..__.___v.
.O actions cannot be made within the allowable out-of-service time. .
m
-v- . .
0~ l. 3/4.4.2 and 3/4.4.3 SAFETY VALVES _ - 1 The pressurizer code safety valves operate to preventEach the RCS safety from being pressurized above its Safety Limit of (2750) psig. i [; valve is designed to relieve Ibs per hour of saturated steam at the - !
' valve set point. f The relief capacity of a single safety valve is adequate toInrelieve the - any overpressure condition which could occur during shutdnwn.
event that no safety valves are OPERABLE, an operating DHR loop, con-nected to the RCS, provides overpressure relief capability . - and will prevent RC5 overpressurization. During operation, all pressurizer code safety valves must be OPE?.ABLE to prevent the RCS from beir.g pressurized above its safety limit of (2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient. , g ( j . , I i l I o i .
. . - - . : ~ . . ._ .
May 7, 1975 1 B&W-STS' S 3/4 4-1 t -
a. ,
E
f
'i -u * .s . -
3/4.7 PLAffT SYSTEMS I
) ' . BASES .-
3/4.7.1 TURBIf!E CYCLE _ i 3/4.7.1.1 SAFETY VALVES _
.l l s ensures . The' 0PERABILITY of the main steam line code safety va ve j that the secondary system pressure will be limited to within its j .
pressure of ( The maximum relieving capacity is asso transient. trip from 100!; RATED THERMAL F0WER coincident with a . i condenser heat sink (i.e., no steam bypass to the condenser).
,j i The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME B Pressure Code,1971 Edition. The total relieving capacity (for all ) percent of the total secondary steam flow of ( ))lbs/hr which islbs/hr at 10 valves on all of the steam lines is ( -l '
that sufficient relieving capacity is available in POWER. safety valve failure. h
' , ~ ~~
i STARTUP and/or POWER OPERATIOil is allowable with safe inoperable within the limitations of the ACTIO:1 requirement i i of the reduction in secondary 'ings ofcvstem the 4'=k.a steam flowchannels.
;C.u ,;dbr and THERMA by the reduced reactor trip s 'the following bases:
The reactor trip setpoint red u ions are derived on 3
- SP = (X) - (Y) V) -x (105.5) .g i
where: i i I yk, fb;d SP = reduced Weir % power trip setpoint in percent of '
; RATED titer'4AL POWER ,
l V = maximum number of inoperable safety valves per steam lin - \ (1,05.5) H4 = Nc'W FMOva *pewer trip setpoint sp.ecified in Table r ~ X = Total relieving capacity of all safety valves per steam
. line in lbs/ hour '
Y = Maximum relieving capacity of any one safety valve in
*f ' lbs/ hour .
5
~~ ~ ~ ~ " '
t- . - . . . .. '
',May 7, 1975 J _ B 3/4 7-1 B&W-STS, t . <
j .. :. .
. 9 CDecee t.}}

ML19329A993

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