ML19317F536

Tech Specs 3 & 4 for RCS

ML19317F536
Person / Time
Site:	Davis Besse
Issue date:	08/13/1976
From:	TOLEDO EDISON CO.
To:
References
NUDOCS 8001150850
Download: ML19317F536 (123)

Text

_.

w 50-346 Davis Bess;e Unit 1 Technical Spec i f ic at i ons with Previously Accepted Changes, w/ltr DTD. 8-13-76 NOTICE -

THE ATT ACHE D FILES ARE OFFICI AL RECORDS OF THE DIVISION OF DOCUME NT CONT ROL THE Y HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD AND MUST BE RETURNED TO THE R E.C O R DS F_ACJ L I T Y BRANCH 016 PLE ASE DO NOT SEND DOC UM E N TS CHARGED OUT THROUGH THE M All REMOV AL OF ANY PAGE tSi F ROM DOCUMENT FOR REPRODUC TION MUST l BE RE FERRED TO FILE PERSGNNE L l

1 DE AOLINE RE TURN DA TE _ _ _ _ _

l

{

1 i -

I RECORDS F ACILITY BR ANCH soc 1:  ;$50 g

s

• a

. THIS PAGE CrEN ,yN;::.;3 RECEIPT OF INFORt4. r:O;j i:p' i ,' - Af.PLICANT

.. REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTE!! LEAXAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTE'45 The RCS leakage detection systems required by this specification are

. provided to detect and monitor leakage frcm the Reactor Coolant Pressure

=

Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Bcundary Leakage '

Detection Systems", May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE . . . ,

PRESSURE SOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure, boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTC0Wil.

Industry experience has shown that, while a limited amount of leakage is expected from the RCS, the UNIDENTIFIED LEAKAGE portion of this can be reduced to a threshold value of less than 1 GPM. This threshold value is suffit.1ently low to ensure early detection of additional leakage.

The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contributicn from tube leakage will

. be limited to a small fraction of Part 100 limits in the event of either a steam generater tube rupture or steam line break. The 1 GPM limit is

. consistent with the assumptions used in the analysis cf these accidents.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allcwance for a

limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. -

The CORTROLLE LEAKAGE limit of) (go GPM restricts operation with a

, total RCS leakage all RC pump seals in excess of ( )GPM.

l0 N. ___

0

• IH

'3

~

y . k 3 DAVIS-BESSE, UNIT 1 B 3/4 4-4 JE 2 2 g 4 . .

I

. l 1

1 1

l l

THIS PAGE WI PEN 2!NG RECEIPT OF REACTOR C00LANT SYSTEM INFOR.hi!ON FROM InE AFFLICANT

(~

BASES The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the

. vessel becomes tne centro 111ng location. The thermal gradients estab-lished during heatup prcduce tensile stiesses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lcwer bound ,

curve similar to that described for the heatup of the inner wall cannot

,be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of

- interest must be analyzed on an individual basis. ,

The heatup limit curves, Figures 3.4-2 and 3.4-3, are composite curves which were prepared by detennining the most conservative. case, with either the inside or cutside wall controlling, for any heatup. rate up to 100*F per hour. The cooldown limit curves, Figures 3.4-2 and

(, '

(3.4-31 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldewn thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of c EFFY, S~ .

The reactor vessel materials have been tested to determine their initial RT,  ; the results of these tests are shcwn in BASES Table 4-1.

Reactor opdhItion and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference tem-perature,baseduponthefYd![c.e and copper content of the material in question, can be predicted using BASES. Figures 4-1 and 4-2. The heatup

. and cooldewn limit curves, of Figures 3.4-2 and 3.4-3 inclid e p . 8 at the end of 2 c.rPY, as well

dicted adjustments for this shift in RT. Stas adjustments for possible errors instruments.

The actual shift in RT., of the vessel material will be establishe'd periodically during operatiBS by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside the radius are essentially identical, the measured tran-sition shift for a sample can applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be DAVIS-SESSE, UNIT 1 B'3/4 4-10 JUL 2 21976 s .  !

i

. EMERGENCY CORE COOLING SYSTEMS BASES _

3/4.5.2 and 3/4.5.3 ECCS SL'BSYSTEMS The OPERABILITY of two independent ECCS subsystems with RCS average temperature > 2FO'F ensures that sufficient emergency core cooling

- capability wIll be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem

~

operating in conjunction with the core flooding tanks is capable of supplying sufficient core cooling to maintain the peak cladding tempera-tures within acceptable limits for all postulated break sizes ranging 2 from the double ended break of the largest RCS cold leg pipe downvard. -

In addition, each ECCS subsystem provides long term core cooling capability in the recirculation r..ade during the accident recovery period.

With the RCS temperature below 280 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. _ , ,

The Surveillance Requirements provided to ensure OPERABILITY of

( each component ensures, that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPEPABILITY is maintained.

-i'cwar b rcair:d t; b; r:mncd fr;m :n; v .M .*kh fai% t an

-eiftgh f;G.r: cri t ria. The decay heat removal system leak rate surveilk. ice requirements assure that the leakage rates assumed for the systen auring the recirculation pnase of the low pressure injection will not be exceeded. l 3/4.5.4 80 RATED UATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on BWST minimum volume and boren concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condi-tion following mixing of the BU5T and the RCS water volumes with all control rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not' usable because of tank discharge line location or other physical l characteristics. The limits on contained water volume, and boron concentration ensure a pH value of betweca (8.5) and (11.0) of the solu-tion recirculated within containment after a design basis accident. The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mecharical systems and components.

DAVIS-BESSE, UNIT 1 B 3/4 5-2 JUL 16 iyt; 4 .

l CONTAff; MENT SYSTEMS BASES -

3/4.6.1.4 INTERNAL PRESSURE ,

The limitations on conta'inment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the ccntainment peak pressure does not exceed the design pressure of 40 psig' during LOCA conditions. ,

' The maximum peak pressure obta'ned from a LOCA event is 37 psig. .

The limit of 1 psig for initial positive containment pressure will limit the total pressure to 38 psig which is less than the design

. pressure and is consistent with the safety analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that g

the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the contain-ment steel vessel will be maintained comparable to the criginal design 3g standards for the life of the facility. Structural integrity is requi to ensure that the vessel will withstand the maximum pressure of psig -

- in the event of a LOCA. A visual inspection in conjunction with ype A leakage tests is sufficient to demonstrate this capability.

3/4.6.2 DEPRESSURIZATIONANDC00LIfd' SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SVSTEM The OPERABILITY of the containment spray system ensures that contain- )

ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reducticn and resuitant lower contai.nment l

l

~

. DAVIS-CESSE, UNIT 1 B 3/4 6-2 JUL 3 01976 j

CO ITAlt:MEf!T SYSTEMS BASES- __

3/4.6.5 CCMBUSTIBLE GAS C0:ITROL '

The OPERD' ITY of the equipment and system equired for the detection and cont of hydrogen gas ensures .at this equipment will be available to mainta' the ring hydrogen con tration within containment below its flammaole limit post- sA conditions. Either recombiner unit (or the purge system) is 'a of controlling the expected hydrogen ium-water reactions, 2) radiolytic generation associated with 1) z' -corrosi of me'tals within containment.

dacomposition of water and 3 These hydrogen control sv . ems are consi ont with the recommendations of Regulatory Guide 1 ~, " Control of CombusJble Gas Concentrations in '

Containment Follos g a LOCA", March 1971.

{eA- -

. 3/4.6.8 SEC0!!DARY C0:1 Tall MEllT_

3/4.6.8.1 EMERGEi:CY VEilTILATIO!i SYSTEM The OPERABILITY of the emergency ventilation ./ stems ensures that containment 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 atmasphere. 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 CPR 100 during LOCA conditions.

.r

~

DAVIS-BESSE, UlIT 1 B 3/4 6-4 . JUL 30 1970

• w e.

F

~

BASES 3/4.6.5 COM3USTIELE CAS CONTROL The OPERABILITY.of the Hydrogen Analy:crs, Containment Hydrogen Dilution System, 7-Containment Recirculation System and Hydrogen Purgc 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.

The two redundant Hydrogen Analyzers determine the content of hydrogen within the -

containment vessel. -

The Containment Hydrogen Dilution (CHD) System consists of two full capacity, redundant, rotary, positive displacement type blowers to supply air to the containment. The CHD System controls the hydrogen concentration by the addition of air to the containme nt 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 fans 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 EEPA filter and charcoal filter prior to discharge to the station vent.

f .

P 6

9% ,

\.

i)+:.,

THIS PA35 T

. MGPT OF E /s. PLICANT

. INFOR.w T.O . . .. : .

PLANT SYSTEMS

/

BASES 3/4 ~  ? AUXILIARY FEEDWATER SYSTEMS

~

The OPERABILITY of'the auxiliary feedwater systems ensures that the Reactor Coolant System can be cooled down to less than 280*F frca normal operating conditions in the event of a total loss of offsite power.

Each steam driven auxiliar, feedwater pump is ca ab of delivering 2

a. total feedwater ficw of <0 gpm at a pressure of (1123 psig to the entrance of the steam generators. This capacity is surricient to ensure -

that adequate feedwater ficw is available to remove decay heat and reduce

. 'the Reactor Coolant System temperature to less than N where the Decay

. Heat Removal System may be placed into, operation. .go'p: ,

3/4.7.1.3 CONDE?! SATE STORAGE FACILITIES .

The OPERABILITY of the condensate storace tank with the minimum water volume ensure's that sufficient water is available for coolde;an of

(

the Reactor Ccolant System to less than 220*F in the event of a total loss of offsite power or of the main feedwater' system. The minimum water volume is sufficient to maintain the P.CS at HOT STAT;DBY conditions for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with steam discharge to atmosphere concurrent with loss of offsite pcaer. The contained water volume limit includes an allowance for water not usable because of. tank discharge line location or other physical characteristics.

, 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These '

values are consistent.with the assumptions used in the safety analyse,s.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERASILITY of the main steam line isolation valves ensures that no more than one steam generator will bicwdown in the event of a steam line rupture. This restriction is required to 1) minimi:e the I

DAVIS-BESSE, UNIT 1 B 3/4 7-2

. . JUL 2 2 197c.

. 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 GRCUP HEIGHT INSERTICN AND POWER DISTRIEUTION LIMITS

. M0 drana ..

This special test exception permits individual control rod to be positioned outside of their specified group heights and in::r . n limits and tn be assigned to other than specified control rod groups, and permits AXIAL POWER IMSALANCE and QUADRANT POWER TILT limits to be exceeded during the performance of such PHYSICS TESTS as those required to 1). measure control red worth, 2) determine the reactor stability 2 index and damping factor under xenon oscillation conditions and 3} -

calibrate AX1AL POWER IMBALANCE and QUADRANT PCWER TILT instrumentation.

' ~

3/4.10.2 PHYSICS TESTS ,

This special test exception permits FHYSICS TESTS to be perfcrmed at less than or equal to S% of RATED THERMAL PO'JER and is required to verify the fundamental nuclear characteristics of the reactor ccre and related instrumentation.

( '.

3/4.10.3 NO FLO'J TESTS - .

.. This special test exception permits reactor criticality.under no flow conditions and is required in order to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 SHUTCC',lN MARGIN This special test exception provides that a minimum amount of con-trol rod worth is immediately.available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

DAVIS-BESSE, UNIT 1 B 3/4 10-1 JUL 22 ny5

' ~

[

d .

1 1

l 4

('

f .

]

THIS PAGE OFEN FT-M . 'G RECEIPT OF INFORI.t;-;Ticii E R ,,, , E APPLICANT un -y

( . . .

i .

! EXCLUSI0ft AREA

~

FIGURE 5.1-1 .

4

.I i -

t DAVIS-BESSE, UllIT 1 5-2~

- JUL 01 n. .

9 4

= eme e ,

w-w- ,w.y- ,- e . 9,,., y g + p, m

,~-

, , _w c-. r . , _ , ,-_,c__ -- ,e+

i i

t

~

THIS PAGE OPEN PEN b~'^ RECElPT OF INFORMATION' g FROs 1.E AFPLICANT

( .

LOW POPULATIO!! 20f E FIGURE 5.1-2 DAVIS-BESSE, UNIT 1 5-3 JUL 01197a l

i

. -- . . .. . ......l l

9

.i THIS PAGE TEN rFN ~..NG RECEIPT OF m f r 'M INE APPLICANT DESIGN FEATURES

( .

5.4 REACTOR COOLANT SYSTEli

• DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:-

O

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to applicable Surveillance Requirements.

b. For a pressure of 2500 psig, and ~

a

' c. For a temperature of 650'F, except for the pressurizer and -

pressurizer surge line which is 670*.F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,110 + 200 cubic feet at a nominal T,yg of 525'F.

5.5.1 The meteorological tower shall be located as shown on Figure 7' -

5.1,1. ,

5.6 FUEL STORAGE ,

CRITICALITY .

5.6.1 The new and spent fuel storage racks are designed and shall be maintained with a nominal 7tg inch center-to-center distance between fuel

.assembliesplac'edinthestoragerackstoensureak[,equivalentto

< 0.95 with the storage pool filled with unborated wl 6r. The K of 7 0.95 includes a conservative allowance of 15 ak/k for uncertai8[fes as described in Section (Pr(1 of the FSAR.

L. I DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation .

DAVIS-BESSE, UNIT 1 5-5 JUL 01 1975

• + e mp p g , e m #

r

f THIS PA'M ' N pcx . :3 RECE!PT OF i

DESIGil FEATURES 1 FROea i..E APPLICANT

( ~~

CAPACITY .

S.6.3 The fuel storage pool is designed and shall be maintained with a storage capacity limited to more than gAQ fuel assemblies. -

-l l

5.7 COMP 0lE!!T CYCLIC OR TRA!1SIEfiT LIMIT l 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limit of. Table 5.7-1.  ;

. *l 4  ;

1 l

i

( - - .

DAVIS-BESSE, UtlIT 1 5-6 JE 01 'm

2

(- . .

4 1 A 9

.f .

J

( .

J THIS PAGE OPEN PEN. iriG RECE:PT OF

INFORA6T!ON FROi.\ ThE Ar'Fl.lCANT 1

A10

=

l .

i i

l -

f 1 .

f 0FFSITE ORGANIZATION i

Figure 6.2-1 JUL 2 0~ ,5., -

6-2 O

, DAVIS-BESSE,tJNIT 1

'e .

. . -. - .- . .. - - . _ . - . . .. .- .._. .- . - - - - - --. -- ll

'A

- i

(- ! '

i. -

THIS PA03 Or.E!.: p5,e.: .;G RECE!?T OF

I N F O R m .. C O ,- Facf. s ThE ArPUCANT t -

alO .

n 1

i .

s FACIl.ITY ORGANIZATION Figure 6.2-2 .

I l DAVIS-BESSE, UNIT 1 6 JUl. 2 9 r.

l l

l . <

l 4s- .

t.  ;

a l

l ACMINISTRATIVE CONTROLS C -

6.3 STATION STAFF QUALIFICATICNS 6.3.1 Each member of the station staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, ex::pt f;. the Mee+th rhy;ici;t h : hall :::t er ::::::d t': ^ '"-t^-- ' 'rac'rtrne Cu:d; i. , ::pt:.f:: ':7 . .

6.4 TRAINING .

6.4.1 A retraining and replacement training program for the station staff shall be maintained under the direction of the Training Coordinator .

and shall r.eet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

~6.5 REVIEb' A.' D AUDIT 6.5.1 STATION REVIEW BOARD (SRB)

FUNCTION

(

6.5.1.1 The Station Review Board (SRB) shall function to advise the Station Superintendent on all' matters related to nuclear safety.

2 1

DAVIS-BESSE, UNIT 1 6-5 JUL 0 0 G76 i

d

ACMitilSTRATIVE CCNTROLS P

u COMPOSITICN 6.5.2.2 The Ccmpany Nuclear Review Boaru shall be ccmposed of the:

Chairman: Vice President, FaciTities Development Member: Vice President, Energy Supply Member: General Superintendent, Power Engineering .

and Construction Member: General Superintendent, Transmission and Substations Member: Superintendent, Davis-Besse Staticn Member: Superintendent', Heavy Maintenance Member: Superintendent, i

-- '~

n 6 pe.rg y $6/VIC Member: Nuclear Engineer, Power Engineering M:tmber: Manager, Quality Assurancs ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis; however, no more than two alter-nates shall participate as voting members in CNRB activities at any one

(, time. .

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

MEETING FRE00ENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of unit operation folicwing fuel loading and at least once per six months thereafter.

CUORUM 6.5.2.6 A qu'orum of' CNR3 shall consist of the Chairman or his

. designated alternate and five of the CNRB members including alternates.  ;

No more than a mincrity of the quorum shall have line responsibility for operation of the facility.

i l

DAVIS-BESSE, UNIT 1 6-9

,td 2 2 1975 1

I

m THIS PAGE 52. : . 3 E?CE!?T OF

^

INFOR.h r.C; U... . 2 A.-PL! CANT PLANT SYSTC'S

. SU'lVEILLA:CE eEr1T?E"C:TS (c. ntinrM)

2. Verifyirg within 31 days after rer. oval a laboratory anclysis cf ct le:s: t..c carh a s rt?les d2.. nstr.t2 a renoval efficiency of > 905 for radioactive cathyl iodir'c when t';e s::ples are tested in acccedanca uith A:;SI !!510-1975 (130*C, 9E R.H.) and tha sa..ples prarared by either:

a) Enptying one entira bed frca a re:7.ed adcorber tray, mixing the ad:orbent thorou @iy, and cbtaining .

samples at least two inches in dictater and with a length equal to the thickhsss of the bcd. .

b) Emptyinc a lancitudinal sarple fron an adsorber tray, nixing tt.e adscrbent thorcuchly, and obtcining samples at least ti;o c inchas in diamatar and with a -

length equal to the thickness of the bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demon-strated OPERABLE by alsc: '

a) Verifying that the charcoal adsorbers rcmove > 99%

of a halogenated hydrocarben refrigerant test gas u'en r they are tested in-place in accordance with AliSI :;510-1975 while cperating the systca at a ficw rate of 3300 cfm i 10%, and b) Verifying that the HEPA filtar banks remove > 99%

of the DOP when tney are tested in-place in accord-ance with A:'SI i:510-1975 while operating the system '

at a flow rate of cfm i 10%.

d. At least once per 18 coaths by: q,g

1. Verifying that the pressure drop across he combined HEPA filters cad charcoal adsorber banks is < l'f inches iater Gauge while operating the ventilation system at a flow rate of 3300 cfm i 10 percent.

2. Verifying that the control reca nor*al ventilation system is isolated by a SFAS test signal, Control Roca Venti-lation Air Intake Chlorine Ccncentration - Hich test signal, and a Station Vent Radiation High test signal.

08 b78 DAVIS-BESSE, U::IT 1 ,3/4 7-19 2 . I

G F.EC5iPT Or c TH!S Po^E OP&3i::3 PLAtlT SYSTEMS SURVEILLA' ICE P.E00IPEPEilTS (Continued) .

3. Verifying that the systen maintains the control recr' at a positive pressure of > 0.25 inches RG. relative to the

~ ~

outside atmosphere at a system flow rate of3?oo cfm

+ 105,

e. After each ccmplete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the D0P when they are test in-place ';n accordance wiTh AftSI ~

N510-1975 while operating the filter systsm at a flow rate of J, loo cfm i 10%. .

f. After each ccmplete or partial replacement of a charcoal ad-sorber bank.by verifying that the charcoal adsorbers remove '

> 99% of a halecenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with AtiSI T1510-1975 .

While operating the filter system at a ficw rate ofecfm i 10%.

. l I

l l

1 l

DAVIS-BESSE. UNIT I 3/4 7-20 W

f d ,

. 1 PLM!T SYSTE!!S H'[AULIC ST:UBSERS (CenH nued) r c,t..

.nos c.

• = I l_ L *

• m** 7

• )

..*= _rte :.I n. :.=.*.

• _.* '. .'_=. q ( m" . . s q._o o.. ., . . 4 4.7.7.1.3 At least cnce par i8 manths i. c., :: ":

• a rg et ..: ::ar.i ve sa:.ple of at least 10 hy:*r:;lic saubscrs or at - :est 103 of al: snu: bars listed in Table 3.7-4, '.; hic :ever is less, siull be saicctcd am far.:-

tionally tested to verify . . ..t pistr.n movc?nt, lock up ::.d Piced.

SnuSbe-s greater than 50,000 lbs capacity my '-* excluded fica func- "

tienal testing requirements. Snub'ars celec:ec fer functiona' e :ing shall be selected on a rotating basis. Snubbers identified' ic. Tau;e c 3.7-4 as either "Especially Difficult to Rsm:je" or in "High Ridiatian

~

Zones" may be exe:apted from functional testing provided these scubbses were demonstrated CPERAGLE during previous functional tests. S.W _-..,.. :r-r .- .: ., . ..:. L .,,.. ... ~..._., .

.. . _: .n.. .. .. ~ ..

......,7....

3 m; . ,.. a . .. .__m..'..., _;; cati:a. For ca:h snubber found irone nble

- during these functional tests, :n additional minimum of 10% of cil snubbers or 10 snubbers, whichever is less shall also be functionell.v

. tested until no more failures are .found or all snubbers have bem functionally tested. .

M l .

t t

i .

t

( .

l l

DAVIS-BESSE, U: LIT 1 3/4 7-22 JUL 0 3 13- 3 1

4 .

l

4 8  :

~

- TABLE 4,7 -

Mf6Ty ptmrdfDRAULIC SNUBBER IllSPECTION SCHEDULE ,

Y' os O NUMBER OF SNUBBERS FOUND Ifl0PERABLE NEXT REQUIRED

.El DURING INSPECTION OR DURING IllSPECTI0fl IllTERVAL* INSPECTION ',TERVAL**

N E

M 0 18 mont! , + 25% .

~ l 12 montns T 25%

2 -

6 months T 25%

3 or 4 . 124 days T 25%

5, 6, or 7 62 days T 25%

2.8 31 days T_ 25%

M Y

Y - .

• Snubbers may be cacegorized into two groups, "acce'ssible" and " inaccessible". This categorization shall be based upon the snubber's accessibility for ipspection during reactor operation. These two groups may be inspected independently according to the above schedule. 8

• • The required inspection interval shall not be lengthened more than one step at a time.

I e

e

• g

. g

'N

_ _ _ _ _ _h ___ ________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ -_ _ _ _ .. c, . ,

t.

ELECTRICAL P0'JER SYSTEMS - -

D.C. DISTRIBUTI0il - SHUTDOWN .

7~- ,

LIMITIt:G C0tIDITIO!1 FOR OPERATI0ft 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:

1 - 250/125-volt D.C. MCC, and

- 2 - 125-volt battery banks and chargers sypplying the above

. D.C. MCC. ,

APPLICABILITY: MODES 5 and 6. ,

ACTION:

With less than the above complement of D.C. equipment and bus OPERABLE, establish C0:iTAlliMENT INTECP'TY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMEilTS 4.8.2.4.1 The above required 250/125-volt D.C. MCC shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 125-volt battery M and charger shall be demcnstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

DAVIS-BESSE, UNIT 1 3/4 8-10 J31 t : ::: '

7i

TH!S p/ 17 .- :.. :.y . ,,

3/4.9 REFUELING OPERATIONS BORON' CONCENTRATION i..E APPLICAN O

LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concen-tration of all filled portions of the Reactor Coolant System and the re-fueling canal shall be maintained uniform and sufficient to ensure that the more restric.tive of the following reactivity conditions is met:

a. Either a Keff of 0.95 or less, which includes a 1% ak/k conservative allowance for uncertainties, or

• 1ir00

. b. A boron concentration of > ppm,.which includes a 50 ppm conservative allowance foF uncertainties.

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 10 gpm of 8750 ppm boric acid solution or its equivalent until K u is Feduced to < 0.95 or the boron concentration is restored to >M ppm, whicheveF is the more restrictive. The provisions of Specification,3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any safety or regulating rod in excess of 3 feet from its fully inserted position.

4.9.1.2 The boron concentration of the reactor pressure vessel and the refueling canal shall be determined by chemical analysis at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ,

• ThereactorshallbemaintainedinMODE6shenthereactorvesselhead is unboited or removed.

DAVIS-BESSE, UNIT 1 3/4 9-1 . JUL 01 19/6.,

1

i REFUELIt:3 OPERATIONS .

( ItiSTRUMENTATIOfi .

LIMITI!;G CONDITIO'i FOR OPERATION C

3.9.2 As a mininte, two source range neutron flux monitors shall be operating, each with contir.ucus visual indication in the control rocm, and en: ith :.dit': '-f':::i: '- th: ::nt:f :n: :nd : ntrol rcer.

. APPLICASILITY: MODE 6. ,

ACTICN: .

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.

/

SURVEILLANCE RECUIREMENTS 4.9.2 Each scurce range neutron flux monitor shall be demonstrated OPERABLE by perfornance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of CORE ALTERATIONS, and

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.

DAVIS-BESSE, UNIT 1 3/4 9-2 JUL 01 197E LJ

REFUELIf;G OPERATIO!!S -

C0:lTAIt: merit PE!iETPATIO!iS LIMITI :G CC4;DITIO 1 FOR OPEPATIO:1 ,_

3.9.4 The containment penetrations shall be in the following stttus:

a. The equipment door closed and held in place by a minim.m of four bolts.

b. A minimum of one door in each airlock closed, and .

c. Each penetration providing direct access from the containment atmosphere to tha outside atmosphere shall be either:

1. Closed by an isolation valve, blind flange, or manual valve, or

2. Be capable of being closed by an OPERABLE automatic containmen purge and exhaust isolation valve.

• M A J. 6 - J1 '

~

APPLICABILITY: During CORE ALTERATI0!is or movement of irradiated fuel

. within the containment.

ACTIO!!: .

With the requirements of the above specification not satisfied, in,;cdiately suspend all operations involving CORE ALTERATICris or movement of irradicted fuel in the containment. The provisions of Specification 3.0.0 cre not applicable.

SURVEILLAf;CE REQUIREMEllTS

[

4.9.4 Each of the above required containment penetrctions shall be determined to be either in its closed / isolated condition or capable of being closed by an CPERABLE automatic contairment purge and exhaus:

valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once por 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment by:

. a. Verifying the penetrations are in their isolated condition, or

b. . Testing the containment purge and exhaust valves per the applicable portions of Specification.4.6.3.1.2.

DAVIS-BESSE UNIT 1 3/4 9-4 pUL 01 in:

- L

.1 4

' ~

REFUELIt CPERATIONS .

(,

_ FUEL HANCLI?;G ERICGE OPERABILITY LIMITING CC"DIT!C" FCR CFEP; TIC" C

3.9.6 The fuel handling bridges shall be used for movement of control rods or fuel assemblies and shall be OPERA 5LE with:

a. A hoist minimum capacity of 3250 pcunds.,and ,

e . a

b. Ahoistoverloadcutofflimitopg750 pounds. -

APPLICABILITY: During movement of control rods or fuel assemblies within tne react.or pressure vessel.

ACTION: ,

With the requirements for bridge OPERABILITY not satisfied, suspend use of any inoperable bridge frem operationc involving the movement i - of control reds or fuel assemblies within the reactor pressure vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLA"CE REOUIREMENTS 4.9.6 Each fuel handling bridge used for movement of control rods or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> 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 lead cutoff when the hoist load exceeds

. 2750 pounds.

x DAVIS-BESSE, UNIT 1 3/4 9-6 Jill 0 " 576

_ s_

, l 1

REFUELIf:G OPEPATICf;S TH;T PA. '5 . SEN .:'-c!' " -3 RECE:PT OF '

INFORt.. h 3.s FR /.a i;J AFPUCANT '

CRA!;E TRAVEL - FUEL HANDLI!;G BUILDIf;G p

LIMITIflG CONDITION FOR OPERATIO!i 3.9.7 Loads in excess of pounds shall be prohibited from travel over fuel assemblies in the storage pool, a W

APPLICABILITY: With fuel assemblies and water in the s.t.eeege pool.

ACTION. j With' the requirements of the above specification not satisfied, place .

the crane load in a safe condition. The privisions of Specification

. 3.0.3 are not applicable.

SURVEILLAriCE RE0VIREMENTS l 4.9.7 Crane interlocks and/or physical stops which' prevent crane travel with loads in excess of pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane operation and at least once per 7 days during crane operation.

4 -

p S

e

• DAVIS-BESSE, UNIT 1 3/4 9-7 ]

JUL J . .s,. l l

r-l ll

1

~

~

REFUELItiG OPERATIO!lS , ..

COOLANT CIRCULATIOil LIMITIf;G C0ilDITITl FOR OPERATI0ft 3.9.8 At least one decay heat removal loop shall be k ;;;r;tka.

~

APPLICABILITY: MODE 6.

. ACT10ti: -

a. With less than one decay heat removal locp in operation, except as ,

provided in b belcu, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. .

b. The decay heat removal loop may be removed from operation for up to one hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of CORE ALTERATI0llS in the vicinity of the reactor pressure vessel (hot) legs. -

c. The provisions of Specification 3.Q.3 are not applicable.

SURVEILLAftCE REOUIREME?tTS 4.9.8 At least one decay heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of > 2S00 gpm at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A k

DAVIS-BESSE, Ut(IT 1 3/4 9-8

. JUL 01 1976 1-

i

~ -

REFUELI!is OPERATIO.*iS .

g- CONTAI""E.*iT PURSE A'iD EXHAUST ISCLATIO1 SYSTEM ,,

s -

O LIMITI"3 C0!i?ITI0'! FOR OPERATION 3.9.9 The containment purge and exhaust isolation system shall be OPERABLE.

APPLICABILITY: MODE 6. -

ACTION- -

Wit 6 the containment purge and exhaust isolation system inoperable, -

close each of the purge and exhaust penetrations providing direct access frca the containment atmtsphere to the outside atmosphere.

The provisions of Spec,ification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS -

4.9.9 The ccntaf rnent purge and exhaust isolation system shall be demonstrateo OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> crior to the start of and at

. least once per 7 days during CORE ALTE1*.TIOnS by verifying that contain-ment purge and exhaust isolation occurs on manual initieuun anu usi e high radiation test signal from ..... ..' the SFAS  ;.m.. .- c..;tien

.;na;

: .

\

DAVIS-BESSE, UNIT 1 3/4 9-9

, JUL 01 ic76 s

. l THIS PA'3E OPEN pt-N .;,.lG RE.~E!PT OF INFORM...T.ON FR3,.1 T..E AFPLICANT 3/4.1 REACTIVITY C0!! TROL SYSTE.5

? ~

BASES 3/4.1.1 30?ATIO:1 C0" TROL 3/4.1.1.1 SHUTDOWN MARGIll A sufficient SHUTCO'.N "ARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are centrollable within acceptable limits, and 3) the reactor will be maintained sufficiently  ;

subcritical to preclude inadver: ant criticality in the shutd:wn condition.

Du' ring Modes 1 and 2 the SHUTDC'.N :'ARGIN is kr.cwn to be within limits .

if all control rods are OPEPABLE and withdrawn to or beyond the S: r"cr limits.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS baron concentration and RCS Ta The most restrictive conditica occurs at EOL, with T atno5c,a.d operhting temperature, ;nd i; :n;;ic.t;d with ; p;;t;Med ;tes.... '.ic.s L. esk .;;ident ar.d r;;ultin; a::::ntr:11;d 3:0 cc,cid;wa. I n ;.N s c.E'.s 5 . 5 - ; ." U.i s &;2 6.J.

a mini-.U:. CEUTCCWd IIAECIN ;f (0.00)3 h/k i; i.ii tid $lj . Q.. ad ;;

ccatrol th: r;;:ti . ity tran:icnt. Accordingly,- the SHUTDCWN MARGIN required is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

3/4.1.1.2 BORON DILUTION .

A minimum flow rate of at least 2800 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be

. gradual through the Reactor Coolant System in the core during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2800 GPM will circulate an equivalent Reactor Coolant System

. volume of 12,110 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron concentration reduction vill be within the capability- for operator recognition and control.

3/4.1.1.3 MODEPATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance require-ment for measurement of the MTC each fuel cycle are 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 MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.

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

. Jill, 2g 1976 d

I

REACTIVITY CONTROL SYSTEMS BASES

(' .

3/4.1.2 BORATION SYSTEMS (Continued) stable reactivity c:nditicn of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. ,

The boron capability required below 200*F is sufficient to provide a

toSHUTD0'.!N MARGIN requires 140*F. This condition o,f 1% ak/k after either ( xenon decay)and cooldown frca 200 borated water from the boric accid storage system or ( ) gallons of  :

(1800) ppm borated water from the borated water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical charac-teristics. The limits on contained water volume, and boron concentration ensure a pH value of between (8.5) and (11.0) of the solution recirculated within containment after a design basis', accident. The pH band minimizes the evolution of iodine and minimizes the effect of. chloride and caustic stress corrosion cracking on mechanical , systems and components. -

The OPERABILITY of one boron injection system during REFUELING i

, ensures that this system is available for reactivity control while in MODE 6. -

3/4.1. 3 MOVABLE CONTROL ASSEMBLIES .

The specifications of this s'ection (1) ensure that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTCOWN MARGIN is maintained, and (3) limit the potential effects of a rod -

ejection accident. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure

., compliance with the control rod alignment and irrerti-- limits.

. ' a.

The ACTION statements which perpit limited A.ve(variations

/dueadfrom the basic requirements are acccmpanied by additional restricticas whict 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 exceeding the assumptions used in the safety analysis.

The position of a rod declared inoperable due to misalignment should not be included in computing the average group position for determining the OPERABILITY of rods with lesser misalignments. .

DAVIS-BES(E UNIT 1 B 3/4 1-3

. JUL 30 1975

. .a e

- i

REACTIVITY C0!! TROL SYSTEMS -

BASES p .

3/4.1.2 BORATI0ft SYSTEMS (Continued) stable reactivity ccnditien of the reactor and the additional restrictions prohibiting CORE ALTERATICilS and positive reactivity change in the event the single injection system beccmes inoperable. ,

The baron capability required below 200*F is sufficient to provide a SHUTD0Uti IGRGI:1 o,f 1", t.k/k after xenon decay and cooldcun from 200*F to 140*F. This condition requires either ( ) gallons of (12,250) ppm borated water frca the boric accid storage system or ( ) gallons of 2 (1800) ppm borated water from the barated water storage tank.

The. contained water volume limits include' allowance for water not available because of discharge line location and other physical charac-teristics. The limits on contained water volume, and boron concentration ensure a pH value of between (8.5) and (11.0) of the solution recirculated within containment after a design basis ' accident. The pH band minimizes the evolution of iodine and minimizes the effect of. chloride and caustic stress corrosion cracking on mechanical systems and ccmponents. -

The OPERABILITY of one boron injection system during REFUELIllG ensures that this system is available for reactivity control while in MODE 6. .

3/4.1.3 MOVABLE C0flTROL ASSEMBLIES .

The specifications of this s'ection (1) ensure that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTUOWil MARGIfl is maintained, and (3) limit the potential effects of a rod -

ejection accident. OPERASILITY of the control rod pcsition indicators is required to determine control. rod positions and thereby ensure

., compliance with the control rod alignment and i- ertir limits.

The ACTI0rt statements which peht limi from the basic requirements are accompanied by additional restrictions which ensure that the original criteria a.e. met. For example, misalignment of a safety or.

regulating rod requires a restriction in THERMAL POWEP., The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in the safety analysis.

The position of a rod declared inoperable due to misalignment should not be included in computing the average group position for determining the OPERABILITY of rods with lesser misalignments. .

. DAVIS-BESSE, UtlIT 1 B 3/4 1-3 JUL 3 0 ;975  !

\

l

. ~w-4 .

_i e

. . t

. 3/4.2 POWER DISTRIBUTION LIMITS

([ .

BASES The specifications of this section provide assurance of fuel integrity during Condition I (Nomal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minircum 0.*lBR in the core

> l.32 during nomal operation and during short term transients, (b) maintaining the peak linear power density < 18.4 kw/f t during normal operation, and (c) maintaining the peak poser density < 20.4 kw/ft during short term transients. In addition, the above criteria must be met "

in . order to meet the assumptions used for the loss-of-coolant accidents.

The power-imbalance envelope defined in Figures 3.2-1 and 3.2-2, and the insertion limit curves, Figures 3.1-1 and 3.1-3 are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temoerature will not exceedOperation a LOCA. the Final outside of the Acceptance Criteria of 2200 F following "I:onstitute a situation that would power-imbalance envelope alone does not cause the Final Acceptance Criteria to be exceeded should a LOCA cccur.

The power-imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the centrol rods are at the inser-( tion limits, as defined by Figures 3.1-1 and 3.1-3 and if aQ percent QUADRANT POWER TILT exists. Additional conservatism is int educted by application of:

• 4A*&

a. Nuclear uncertainty faqtors. .

b. Themal calibration uncertainty.

c. Fuel densification effects.

d. Hot rod manufacturing tolerance factors.

• i The ACTION statements which pemit limited variations trom the basic requirements are accompanied by additional restrictions which ensures that the original criteria are met. r The definitions .f the design limit nuclear power peaking factors as i used in these specifications are as follows: ,

F Nuclear Heat Flux Hot Channel Factor, is defined as the maximum 9 local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions. .

l

- l DAVIS-BESSE, UNIT 1 B 3/4 2-1 .

JUL 3 01976

~

4. .

\ .

POWER DICRIBUTION LIM"S_ ,

BASES _

N Huclear Enthalpy Rise Hot Channel Factor, is defined as the F

AH ratio of the integral of linear. pcwer along the rod on which minimum EBR occurs to the average rod power.

It has been determined by extensive analysis of possible oper; ting power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, provided: ,

. F q1 2.94; F gi 1.71 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 be met provided the fol-lowing conditions are -

maintained.

1, Control rods in a single group move together with no individual rod insertion differing by more than + 6.5% (indicated position) from the group average height. .

2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.

3. The regulating rod insertion limits of Specificati6n 3.1.3.6 are maintained.

4. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is . measure of the difference in power between the top and bottem halves of the core. Calculctions of core average axial peaking factors for many plants and measurements from operating plants under a variety of ocerating conditions have been correlated with AXIAL POWER IMSALANCE. The correlation shows that the design power shape is not exceeded if the AXIAL POWER IMSALANCE is maintained between +% percent and - gG'

. percent at RATED THERFAL POWER. Ay The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn

- to minimum allowable control rod insertion and are the core ONBR design basis. Therefore, for operation at a fraction of RATED THERFAL POWER, the design limits are met. Whenusjngincoredetectorstomakepowerdistribu-tion maps to determine F g and F'g:

3 Meas , shall be

a. The measurement of total peaking factor, F increasedby1.4percenttoaccountformabufacturing ulcrances and further increased by 7.5 percent to account for measurement error.

DAVIS-BESSE, UNIT I B 3/4 2-2 N. 2 2 ;gyg

g]

r THis :pfm= s 7%-

A a.:.;4 .,u .

INfoRtis.- T O,4 * ** RECBPT OF

{ POWER DISTRIBUTION LDtITS /4 l..E AFPUCANT j

BASES

b. The measurement of enthalpy rise hot channel factor, FN s

~

- beincreasedby5percenttoaccountformeasurementeNo,r. hall

- For Condition II events, 'the core is protected frem exceeding 20.4 kw/ft locally, and from going below a minimum D iSR of 1.'32, by

. automatic protectio'1 on power, AXIAL POWER D1SALANCE, pressure and .

temperature. Only conditions 1 through 3, above, are mandatory since thh AXIAL POWER IMBALA! ICE is an explicit input to the Reactor Protection -

System. ,

The QUADRAHT POWER TILT limit assures that the radial power distribution satisfies the design values, used in the power capability .

analysis. Radial power distribution megsurements are made during -

startup testing and periodically during power operation.

The QUADRANT POWER TILT limit of at which corrective action is required provides DUB and linear heat ga ration rate protection with x-y i plane power tilts. A limiting tilt o 4.5* can-be tolerated before the margin for uncertainty in F is deplete . he limit of 4% was selected to provide an allowance for0 the uncertainty associated with the indicated power tilt. In the event the tilt is not c Frected, the margin for .

uncertainty on F is reinstated by educing the pcwer by 2 percent for each percent of Silt in excess of 4 3/4.2.5 DNB PARAMETERS N %C) 1 ,

The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout ,

, ;each analyzed transient. .

1 .

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored

. within their limits following load changes and other exoected transient.

operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured ficw such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

l l

. DAVIS-BESSE, UNIT l' B'3/4 2-3 M E ? '375 l

l l

.. - _ , - - - - ._ _ . I1

. c

• J e

opp.n pgpiNMG nECEl?T OF i

-- THIS PAG: " "*APPUCAbb, r

1NFOR>""P'Oc' ROia.ijnT T PHE ^

. u t

S e

i .

g e l

I r

.. INCORE INSTRUMENTATI0ft SPECIFICATION ACCEPTABLE MINIMUM AXIAL Ii!BALa.NCE ARRANGEMENT BASES FIGURE 3-1

. DAVIS-BESSE,UilIT1 B 3/4 3-4' JUL 22 975 e

'=.m I

~

('

s

~

2 i 'THIS PAG 3 OPEN PENDING RECE;FT OF INFO.t M. ~!O. I F'DM T .E A. ?'.:CAi4T NOT $0E .

t .

I

l. .

INCORE INSTRUMEilTATION SPECIFICATICN ACCEPTABLE MINIMUM RADIAL TILT ARRANGEMENT ,

BASES FIGURE 3-2 .

, DAVIS-BESSE, UNIT 1 B 3/4 3-5 JUL 2 21976

. 1

- 4 .

3/4.4 REACTOR COOLANT SYSTEM ,

7 BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.32 during all normal operaticns and anticipated transients. With one reactor coolant pump not in operation in one or both 1:)ps, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL FO'.iER IMBALANCE and the 2 Hyclear Overpower Based on Pump Monitors trip, ensuring tnat the CNER

will be maintained above '. at the maximum pcssible THERMAL POWER -

- for the number of reactor co lant pumps in operation or the local

' quality at the point of min'.um DNBR equal to 22%, whichever is more restrictive. , , ;p . ,'

A single reactor coolant loop provi, des sufficient heat removal capability for removing core decay heat while in HOT liAND3Y; however,

single failure considerations require, placing a DHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

( ,

3/4.4.2 and 3/4.4.3 SAFETY VALVES ,

The pressurizer code safety, valves ~ operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig. Each safety valve is designed to relieve 335,000 lbs per hour of s.aturated steam at the valve's setpoint. .

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdcwn. In the

- event that no safety valves are OPERASLE, an operating CHR loop, con-

, nected to the RCS, provides overpressure reli,ef capability and will

' ~

prevent RCS overpressurization. .

During operation, all pressurizer code safety valves must be OPERABLE 2

to prevent the RCS from being pressurized above its safety limit of

.2750 psig. The combined relief capacity of all of the:e valves is greater than the maximum surge rate resulting from any transient.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in acccrdance with the provisions of Section XI of the ASME Boiler and Pressure Code.

DAVIS-BESSE, UNIT 1 B 3/4 4-1 JUL 2 2197g 4 .

. 1

~

e -

REACTOR COOLANT SYSTEM .

BASES

~

3/4.4.4 pRESSURIZQ . .

A steam bubb',e in the pressuri:er ensurcs that the RCS is not a hydraulically solid system and is capable of acccmmodating pressure surges during operation. The steam bubble also protects the pressurizer

. code safety valves and power operated relief valves against water relief. -

The low level limit is based on providing enough water volume to .

. prevent a pc h .. m L.

- ? or a reactor coolant system low pressure condition that would actuate the Reactor Protection System or the ~

' i . 3 . . . . ,. m Safety Feature Actuation System as a result of a reactor .f~p

. aerem The.higt) level limit is based on providing enough steam volume .

to prevent a pressurizer high level as a', result of any transient.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valtes. -

3/4.4.S STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator

.. tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to '

maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanicel damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is exoected to be operated in a manner such that the

, secondary coolant will be maintained within those enemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these chemistry limits, locali:ed corrosion may likely result in stress cerrosion cracking.

The extent of cracking during plant operation wculd be limited by the limitation of steam generator tube leakage between the primary coolant ~

systcm and the secondary coolant system (primary-to-secondary leakage = 1 GFM).

Cracks having a primary-to-secondary leakage less than this limit'during operation will have an adequate margin of safety to withstand the loads

[

DAVIS-BESSE, UNIT 1 B 3/4 4-2 J[JL 2 2 ;: -

l ..

t .

l

o

' TABLE 3.3 g. .

RADIATION MONITORING INSTRUMENTATION '

Y MINIMUM APPLICABLE ALARM / TRIP MEASUREMENT.

M - CilANNELS '

SETPOINT RANGE ACTION N INSTRUMENT OPERABl.E MODES

-T e 1. AREA MONITORS

=

• a. Fuel Storage Pool Area Emergency Ventilation (1 - 10 5) cpm 15 System Actuation (1)

' ** 1(2) x background

. 2. PROCESS MONITORS

. R a. Fuel Storage Pool Area

1. Gascous Activity -

y Normal Ventilation -

6 O System Isolation 1 .

1 3 x 10-9 pci/cc 10 - 10 cpm 15

11. Particulate Activity -

Normal Ventilation System 6 Isolation 1 1 1 x 10~10 pci/cc 10 - 10 cpm 15

b. Containment

1. Gaseous Activity

-e) Purac 'e-Emhettst- 106epm A x 10'7 pciTcc P- -10 16

, Isolation -fr--

b) RCS Leakage 7

Detection 1 1, 2, 3, & 4 Not Applicable 10 - 10 6cpm 16 kr. '

11. Particulate Activity FA5

a , . . . , , . -_ . . . m M 6 cf ht&tJon P His -: 1 x 10-10 ucifcc 40---10 -spm. 16

.9

• b) RCS Leakage 6 Detection 1 1, 2, 3, & 4 Not Applicable 10 - 10 cp, j4

-: *With irradiated fuel in the storage pool

' **With fuel in the storage pool or building THIS PAGE OPEN ri-N It G RECEIPT OF INFORMATJOs1 FROM li.E Al PLICANI s ___ . - l

TA3tE 3.3-6 (Continued)

~

TABLE !;37ATION ACTIO:: 14 - With the nu .ber of channels CPEPABLE less than required by the P.inie.:: Channels 0? ERA 3LE requirement, comply with the A*TIO requirements of Specification 3.4.6.1.

ACTION 15 - With the number of channels OPERAELE less than required by the Mini::ren Channels OPERASLE requirement, comply with the ACTIC:: requirenents of Specification 3.9.12.

AC,a._,

. -v.i

.v - ...._~...;_....__

n....

_u.__.,_

. ... o....s ....,-,._.u..,_.

. _ .. . . ..,s.. mw b y tk.: "< :.r c.:-- r:;uir= = :, :::;1y

.n-.-. 0 I'_J.2L: __. __ -__ _ - n -

v ,. _ . . . . e . v . . s.s.s.

ms. _m. ..m . . . .

. s.

..a. a. . m_n...... . e.

j

-f i

l 1

1

. DAVIS-BESSE, UNIT 1 3/4 3-28 JUL 0 , . .. , ,

l 4W i

s

• M + h

~

TABLE 4.3-3 RADIATIONMONITORINGINSTRUMENTATIONSURVEILtINCEREQUIREMENTS 5 CHANNEL MODES IN WHICH i

-

• CHAfiNEL CHANNEL FUNCTI0ilAL SURVEILLAriCE

[ CALIBRATION TEST REQUIRED g INSTRUMENT CilECK w

1. AREA MONITORS E Fuel Storage Pool Area q a. .

m Emergency Ventilation System .

Actuation S R M

2. PROCESS MONITORS

a. Fuel Storage Pool Area

. 1. Gaseous Activity -

flornal Ventilation

• System Isolation S R M

$ 11. Particulate Activity -

ttormal Ventilation w

• A, System Isolation S R H .

w

b. Containment ' *

1. Gaseous Activity *

a) n'qn..t Fvhw.t .

4s<n+Lun "s- . -R-- -M- +

b) RCS leakage detection S R 8M 1, 2, 3 & 4 ii. Particulate Activity abParr,c ! Er'musi 4t h i P + + +

g, b) RCS Leakage Detection S R M 1, 2, 3 & 4

(*

• With irradiated fuel in the storage pool

• • With irradiated fuel in, the storage pool or building T

?

r . t

~~

INSTR'I'ENTATION

,s IN:0;E DETECTCES LIMITI*:3 C N :T:CN FCR C EE* TIC';

3.3.3.2 As a minicum, the incore detectors shall be OPERABLE as speci-fied below.

a. For AXIAL POWER IMEALANCE censurements:

-rse ar e M. 4

1. 4344 detectors, 6Aaee in each of 3 strings, shall lie in the same axial plane with I plane in each axial core half. ,

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.

b. For QUADRANT POWER TILT measurements:

1. Two sets of 4 detectors s' hall lie in each core half.

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

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

. axial plane.

2. Detectors in the same plane shall have quarter core .

radial symmetry.

APPLICABILITY: When the incore detection syste'm is used for surveillance of:

a. The AXIAL POWER IM3ALA"CE, or ,

e

b. The QUADRA"T POWER TILT.

ACTION:

With less than the specified minimum incore detector arrangement OPERABLE, do not use incore detector measurements to determine AXIAL POWER IMBALANCE or QUADRA"T POW'ER TILT. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUTRE"I"TS 4.3.3.2 The incore detector system shall be demonstrated QPERABLE:

O DAVIS-3 ESSE, UNIT 1 3/4 3-30 J-) L .-

8 w/b e

b

i

- ~-

THIS PAC-E .- .: 3 RECE!PT r

'^' IN FC F.r.. .

i . E AFMIIZET"0'; TCRI;;3 Il

57RU:'I!;TATIO*:

N c r - ,a , i c MII4IiCM MEASUP.D'ENT INSTRUMENT I!.sTR'.*MEi.is A!;D SE!;5Co. LCC

• TIC';5 rat:GE OPERABLE

1. Triaxial Ti..e-History Accelographs

a. I

b. I

c. I  ;

d. I -

2. Triaxial Peak Accelographs

a. I

- b.

l

c. I

d. .

I

e. ..

I

' ~

3. ' Triaxial Seismic Switches .

a. l*

I b. l*

l

c. l*

~

d. l*  !

s 1

4. Triaxial Response-Spectrum Recorders

a. l*

b. l

c. I l

d. I l

e. 1

f. l

• With reactor control roca indication s

. DAVIS-CESSE, UNIT 1 3/4 3-33 '

.. f t

. t L_

i THIS PA. 2 ~~;" .-

3. RECE'?T C. ?., ~.3-~,

IN FO R/.'.~ h0. c '. . . 1 E A~PL!CANi M er r o< 9 c - e. ; .-. ---

'T:0!; St.'R'.'EILLA';CE REC'RPE:'E';TS CHA!1t;EL CHA.;!iEL CHAT::!EL TUNC rIO.J c.

' INSTR'."*E';TS A!O SE';!*R LC3TIO*;S CHECK CALI*y.'IC'! TEST

1. Triaxial Tir.e-History Accelographs M* R SA a.

M* R SA b.

M* R SA c.

M* R SA d.

' a

2. Triaxial Peak Accelegraphs .

NA R NA a.

NA R NA b.

NA R NA

- c.

d. NA k NA

e. ,

fM E NA

. 3. Triaxial Seismic Switches

'a . .

• • M P. SA

b. . **M R SA

c. **M R SA

d. **M R .SA ,

/

4. Triaxial Response-Spectrum Recorders

a. **M R SA l

b.

IM R SA

c. NA R SA

d. NA R SA

. e. NA R SA

f. HA R SA

• Except seismic trigger >

• • With reacter control room indication

,' DAVIS-BESSE, U::IT 1 3/4 3-34 M 0 7197s

__ [

1 1

• • en TABLE 3.3-8

^

METEOP.0 LOGICAL MONITOR!?;G INSTRUENTATIOfi .

INSTRUMEriT MItilMUM MIri1 MUM LOCATION ACCURACY OPERABLE INSTRUMENT

1. WI!iD SPEED

a. flominal Elev. 612 1 0.5 mph

• 1

b. tiominal Elev. 827 1 0.5 mph

• 1 ,

2. WIfiD DIRECTIO!!

a. tiominal Elev.. 612 1 5' 1

b. Nominal Elev. 827 i 5' 1 3.' AIR TEMPERATURE - DELTA T

a. Nominal Elev.

Jht' ..

10.1' c 1 M

4. 'k ":' :ki . -5E-4 0.i t Starting speed of anemometer shall be < 1 mph. ,

f DAVIS-BESSE, UNIT 1  ?/4.3-36 .p

_, c

) .

TABLE 4.3-5'

. METEOROLOGICAL MONITORING INSTRUMETATION SURVEILLANCE REQUIREMENTS U CilANNEL CHANNEL i Cl!ECK CALIBRATION kw INSTRUMENT .

y 1. WIND SPEED E a. Nominal Eley. 612 D SA

--t '

- b. Nominal Elev. -

827 ,

D SA

2. WIND DIRECTION -

a. . Nominal Elev. 612 D SA

b. Nominal Elcy; 827 .

D SA R .

[ 3. AIR TEMPERATURE - DELTA T b a. Nominal Eley. #37-612 . D SA

-b. ;knaimr14tcy. 02f- -

S A--

g r-o N

e F- . n,

3

, f,.

If;STRU"Ef;TATIO!;

• D h

Q CHLORI!;E DETECTIO!; SYSTEMS

,h& .

LIMITIt:G C0?;0fTIOl; FOR OPERATIO!; 6 3.3.3.7 Two independent chlorine detection systems, with their alarm / trip setpoints adjusted to actuate at a chlorine concentration of 1 S ppm, shall be OPERABLE.

APPLICABILITY: 1, 2, 3 and 4 ACTIO:::

~

a. With less th'an two chlorine detection systems OPERAGLE, within l hour initiate and maintain operatien of the control room emergency ventilation system in the recirculation racde of operation; restore the inoperable detecticn syste.,to ?.' ERA 3LE

~

status within 30 day or be in at least HOT STA:;DEY ttithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDC',;ii within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLAf;CE REQUIREMEt,'TS .

4.3.3.7 Each chlorine detection system shall be demonstrated OPERABLE by performance of a CHAtii;EL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a C'!ANNEL FUtiCTIO!;AL TEST at least once per 31 days, and a CHA!;;;EL CALILRI. TION at least once per 18 months.

l 1

. i

~

DAVIS-BESSE, UNIT 1 3/4 3-44 M 0 71976 A

/

~r l,p *

,' /

l' 3Qhl:.j[r-y

!!iSTRUMEflTAT10:1

, [ .

CilLORI!!E DETECTIC 1 SYSTEM _

{

s

, . , s.

.-- .LitilTit;G C0'!DITIO!! FOR OPERATION /

3.3.3.6 The chlorine detection sf em, with the alarm / trip setpoint adjusted to actuate at a chlorisconcentration of < 5 ppm, shall be 1

. ,% OPERABLE with, as a minimum,[(two,ff OPERABLE chlorine detectors located in the Reactori Control Roon vtmtilation air intake. l . ; , . , ,. , E

. . a. t f@ - /

• d i s .9 /A 'M .

. a- .

.3 , ,,

's '

APPLICABILITY: ALL MODES ,

a ACTIO:1:

'", With the chlorine detection system inoperable, initiate and maintain operation of the control room emergency ventilation system in the recirculation mode-of operation ; restore the chlorine detection system to OPERABLE status within 30 days or be in at least HOT STA!!DBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'.!1 within t:ie next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREME!!TS 4.3.3.6 The chlorine detection system shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHAN lEL Fut!CTIO:lAL TEST at least once per 31 days, and a CHAfiilEL CALIBRATIO:t at least once per 18 months.

4

.i s

\

DAVIS-BESSE, UtlIT 1 3/4 3-41

. *'*o.,

m O

a

.o

. TABLE 3.4-1 .

REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 3

5 j, STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT

, DISSOLVED OXYGEN

• 1 0 10. ppm 1 ,1.00 ppm o

CHLORIDE 1 0 15 ppm i 1.50 ppm FLUORIDE 1 0 15 ppm 1 1 50 ppm Limit not applicable with T,y 1 250*F.

~

DAVIS-BESSE, UNIT 1 3/4 4-18 g I

l -- L

1 EMERGEf;CY CORE COOLING SYSTEMS r

ECCS SUBSYSTEMS - 9=m==netMMF av5 .

LIMITIfG CONDITIO'i FOR 0?ERATION 3.5.2 Two ECCS subsystems shaIl be OPERABLE with each subsystem ,

comprised of:

a.- One OPERABLE high pressure injection (HPI) pump, a

b. One OPERABLE low pressure injection (LPI) pump,

' ~

c. One OPERABLE decay heat cooler, and ~

d. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and automatically transferring suction to the contain. ent t sump on a borated water storage tank low level signal dering tha recirculation phase of operation. .

APPLICABILITY: MODES 1, 2 and 3. *

~

ACTION:

. a. With one ECCS subsystem inoperable, restore uhe inop- atle subsysted to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in h0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. .

b. 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 5.9.2 within 90 days describing'the circumstances of the actuation arid the total accumulated actuation cycles to date. .

DAVIS-SESSE, UNIT 1 3/4 5-3 JUN 3 01976

^

l

i EMERGENCY CCRE C03L!N3 SYSTE"5

.s SL'R'.'EILLANCE RE0L' IRE"E';TS 4.5.? Each ECCS subsystem shall be derenstrated OPERABL'E:

a. At least once per 31 days by verifying that each vab le (canual, p:.<sr c;erated er autcratic) in the flow patr. that is not locked, sealed cr otherwise secured in position, is in its correct pcsition.

b.

/W9 By a visual ins;ecticn which verifies that no Tocse debris a

(rags, trash, clotning, etc.) is present in th2 cor.ta'nq6nt -

which could be transported to the containment / sump and ceuse restriction of the puro suctica during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to

- establishing CONTAINMENT INTEGRITY, and ,

. 2. Of the areas affected within containment at tFe completion i of each containment

• entry when C0!jTAIRMENT IU!EGRiTY is established. ,

c. At least once per 18 months by:

~

. 1. Verifyir.g autenatic isolation and interlock actic.n of the DHR syste- fr:n the Rea: tor Coolant' System when the Reactor Ccolant System pressure is > 280 psig.

~~ .

4' .

w DAVIS-BESSE, UNIT 1 3/4 5-4 ggg, 3 0 E76

.I '

l EMER3ENCY CO?.E C00LIN3 SYSTEP.S

(% ECCS S'J35YSTE"S - ?l25 2-P LIMITING CONDITION FOR OPED.ATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall .

be OPERABLE:

a. One OPERABLE high pressure injection (HPI) pump,

b. One OPERABLE low pressure injection (LPI) pump, ,

1: . One OPERABLE decay heat cooler, and ,

d. An OPERABLE flow path capabiu  ! taking suction frca the borated water storage tank (BWST) and transferring suction-to the containment emergency sump.

APPLICABILITY: MODE 4.

ACTION: -

^

a. With no ECCS subsystem OPERABLE because of the inoscrrhility of either the HPI pump or the ficw path from the ' vorated water storage tank, restore at least one ECCS subsystem tc rr3Er.qstg status within one hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b. With no ECCS subsystem OPERABLE because of the inoperability of either the decay heat cooler or LPI oump, restore at least one ECCS subsystem to OPERABLE status or maintain the Raacter Coolant System T less than 280*F by use of alternate neat removal methods."V9

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.?.2 within 90 days describing the circunstances of the actuation and the total accumulated actuation cycles to date.

SURVEILI1NCE REOUIREMENTS ,

4.5.3 The ECCS subsystems shall be demonstrated OPERASLE per the applicable Surveillance Requirements of 4.5.2.

DAVIS-BESSE, UNIT 1 3/4 5-6

~ r , <e ,

. ,1 l?

t EMERGENCY CORE COOLING SYSTEMS 77p3 p*, ,'E

• ) FEN ?.=N' is3 RECEIPT OF

g. BORATED WATER STCRAGE TA*;K NFO i - r og pay,. I.- E Mpgm LIMITING CONDITION FOR 03ERATION 3.5.4 The borated water storage tank (BWST) shall be OPERABLE with:

a. A contained borated water volume of between 402,500 and

( ) gallons, 730 0

- b. Between 1800 and,EG60' ppm of boron, and ,

a

^ c. A minimum water temperatyre of 35*F ,

APPLICASILITY: MODES 1, 2, 3 and 4.

ACTION: -

With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STAN05Y within

- the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 3G hours.

i ,

SURVEILLANCE REOUIREMENTS 4.5.4 The BWST shall be demonstrated OPERABLE:

a. P least once per 7 days by:

1. Verifying the contained Lorated water volume in the tank, i

2. Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the water temperature when outside air temperature <35'F.

1 DAVIS-BESSE, UNIT 1 3/4 5-7 JW 3 o sg e

I

i EMERGENCY C00,E CCOLING SYSTEMS 77g3 pn, ,-E 3.*EN r,=N ts3 pygg;PT OF

c. B0MTED WATER STORAGE TA*;K ,

NFO:>. Tog pagf,. i I.- E A.;PLICANT i

i LIMITING CONDITION FOR 03ERATION 3.5.4 The borated water storage tank (BWST) shall be OPERABLE with:

a. A contained borated water volume of between 402,503 and

( ) gallons,

,1300

- b. Between 1800 and,EG60' ppm of boron, and .

c. A minimum water temperature of 35'F ,

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: -

With the borated water storage tank inoperable, restore the tan:< to OPERABLE status within one hour or be in at least HOT STAND 3Y within

- the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

i ,

SURVEILLANCE REOUIREMENTS 2' -

4.5.4 The BWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the contained borated water volume in the tank,

2. Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the water temperature when outside air temperature <35*F.

DAVIS-BESSE, UNIT 1 3/4 5-7 Jgg 3 o g;g e

THIS PAGE OPEH P -N.91NG RECEIPT OF INFORW.T.C04 FRJM I;.E APPLICANT C0tiTAINMENT SYSTEMS _ qo7 7 e gj-r.

INTERNAL PRESSURE LIMITING CCNDITION FOR OPERATIO.l 3.6.1.4 Primary containment internal pressure shall be maintained betweenos"s'gand -s" nc m. jM/ .8 pm APPLICABILITY: MODES 1, 2, 3 and 4.

ACT-ION:

With the containment internal pressure outside of the limits above, restore the internal pressure to'within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be '

in at'least HOT STAIID3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in. COLD SHUTDOWN within the folicwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• 9 i -

SURVEILLANCE REOUIREMENTS 4.6.1.4 The primary containment internal pressure shall be detemined to within the limits at least once par 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 1

l 1

. 1 DAVIS-BESSE, UNIT 1- 3/4 6-7 JUL 2 2 97g

! j j

i

~

TABLE 3.6-2

,-- CONTAINMENT ISOLATICN VALVES (Continued)

TESTABLE PENETRATION VALVE DURING PLANT ISOLATION NUMBER NUMBER FUNCTION OPERATICN TIME (seconcsl 30 DH9A Containment Sump Emergency Recirc Line No 71 31 DH9B Containment Sump Emergency Recirc

. Line No 71 32 RC1773A RCS Drain to RC Drain Tank Yes 10 a

- 32 RC17738 RCS Drain to RC Drain Tank Yes 10 ,

35 AFSSS Amil m,r ice '. ets Lir.e Ycs 10 OC ATCOC A silier., ra d,ater Line Yes 10 37 . FN601' Main Feedwater Line No ' ' '15

, 38 FW612 Main Feedwater Line

• No 10 39 MS100 Main Steam Line , No . 10 39 MS107 Main Steam Line Yes 10 39 M5107A Main Steam Line -

Yes 10 39 ICS11A Main Steam Line No 10 i 39 MS375 Main Steam Line Yes 10 39 MS100A Main Steam Line Yes 10 40 MS101 Itain Steam Line ,

No 10 40 MS105 Main Steam Line Yes 10 40 .MS106A dainSteamLine Yes 10 40 ICS11B . Main Steam Line No 10 40 MS394 Main Steam Line Yes 10 40 MS101A Main Steam Line ,

Yes 10 41 RC232 Pressurizer Quench Tank Circulating

Inlet Line Yes 10 42A ' SA201'O Servica Air Supply Line Yes . 10 428 CVS010E Containment Vessel Air Sample Return Yes 10 I

43A IA2011 Instrument Air Supply Line '

No 10 438 CV5011E Containment Vessel Air Sample Raturn Yes 10 44A CF1541 Core F1 cod Tank Fill and N2 Supply Line Yes 10 l s 44B NN235 Pressurizer Quench Tank N2 Supply

! Line Yes 10

. DAVIS-BESSE, UNIT 1 3/4 6-17 JUL 2 0 97s l,

. l TABLE 3.6-2 ,

C CONTAINMEllT ISOLATI0ft VALVES (Continued)

TESTABLE PENETRATION VALVE DURING PLANT ISOLATION NUMBER NUMBER FUNCTIO:t OPERATION TIME (seconds]

i '

748 CV5010D Containment Air Sample Yes 10 74B CV50110 Containment Air Sample Yes 10 74C DH2735 Pressurizer Auxiliary Spray Yes 10 74C DH2736 Pressurizer Auxiliary Spray- Yes 10  ;

B, ., CONTAINMENT PURGE AND EXHAUST ISOLATION

. 33 CV5005 Containment Vess,el Purge Inlet Line Yes 10 33 CV5005 Containment Vessel Purge Inlet Line Yes

~

10 '

34 CV5007 Containment Vessel Purge Outlet Line Yes 10 34 CV5008 Containment Vessel Purge Outlet Line Yes

• 10

~

C. MANUAL -

17 CV343 Containment Vessel Leak Test Inlet Line No N/A 23 SF1 Fuel Transfer Tube No N/A 24 SF2 Fuel Transfer Tube No N/A

• 25 C533 Containment Spray Line Yes N/A

• 25 CSl7 Containment Spray Line Yes N/A

• 25 SA536 Containment Spray Line Yes N/A

. *25 SA532 Containment Spray Line Yes N/A

• 26 CS36 Containment Spray Line Yes . N/A

• 26 CS18 Containment Spray Line Yes N/A

• 26 SA53S Containment Spray Line Yes N/A 29 DH21 Decay Heat Pump Suction Line No N/A

.. 29 DH23 Decay Heat Pump Suction Line No N/A  ;

[*47A CF2A** Core Flood Tank Sample Line Yes N/A

• 47A CF2B** Core Flo'od Tank Sample Line Yes N/A

\ 3s Ap599 Aux reto wi,cp &14E p.: r,g 36 Ap&cg or t. 1 'ri .

,vc ,,,g

\ DAVIS-BESSE, UNIT 1 3/4 6-19 JUL 2 2 1975 d ll

CONTAI!r!ENT SYSTEMS

('- . .

3/4.6.5 SECONDARY C0!TAlf; TENT .

EMERGENCY VEtlTILATIO!! SYSTEM LIMITING CONDITION FOR OPERATION

~

3.6.5.1 Two independent emergency ventilation subsystems shall be OPERABLE. .

APPLICASILITY: MODES 1, 2, 3 and 4.

ACTION:

With one emergency ventilation subsystem inoperable, restore the inoperable

- subsystem to 0?ERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'!N within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

\ .

SURVEILLANCE REOUIRES'ENTS 4.6.5.1 Each cmergency ventilati'an.a=6 system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Initiating from the control room flow throu5h the HEPA filter and charcoal adsorber train and verifying that the train operates for at,least 10-h:cr: ith th. :.._... v.r, M ',I.' /f M -

2. Verifying that 'each ventilation subsystem is aligned to receive electrical pcuer from separate OPERA 5LEs_ _.;_.. ,

busses. .W

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, by:

l DAVIS-BESSE, UNIT 1 3/4 6-22

+ .

i THIS P.vi2 c.,E N i' N 2 :G RECEIPT OF

,m JNFO L..i.c;q p,ujy ).7.,g g CONTAIMMENT SYSTEMS i SURVEILLANCE RECUIREMENTS (Continued) i .

3. Verifying that the filter train starts autentically on any containment isolation, test signal.

4. Verifying that the filter cooling bypass valves can be manually opened. _

5. Verifying that each system produces a negative pressure '

of > (0.25) inches W.G. in the. annulus at a system flow

. rate of go06 cfm i10%, within (1) minute after a start signal. -

. /[ .

[e.s2E, I N D IA M4$ [ M # #*"

i e. After each. complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place.in accordance siith ANSI

, t N510-1975 while operating the filter system at a flow rate of 9000 cfm i 10%.

f. After each complete or partial replacement of a charcoal ad-sorber bank by verifying that the charccal adsorcers remove

> 99% of a halogenated hydrccarbon refrigerant test gas when they are tested in-place in accordance with NISI N510-1975

~

. while operating the filter system at a flow rate of 7000 cfm i10%.

1

. DAVIS-BESSE, UNIT 1 3/4 6-25 Al 2 *a b/O -

ll

i

p. CONTAINME;1T SYSTEMS SHIELD BUILDI!!G I1TEGRITY_

• LIMITING CCt:DITIO!! FOR CPERATI0ft 3.6.6.2 SHIELD BUILDING INTEGRITY shall be maintained. ,

APPLICASILITY: MODES 1, 2, 3 and 4.

ACTION:

Without SHIELD BUILDING INTEGRITY, restore SHIELD BUILDING INTEGRITY within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY' within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

r

.~

( .

SURVEILLANCE REQUIREMENTS .

3.6.6.2 SHIELD BUILDING INTEGRITY shall be demonstrated at least once per 31 days by verifying that airtight doors and the blowout panels listed in Table are closed except when the airti5ht doors are being used for normal nsit entry and exit.

. So b'l .

DAVIS-SESSE, UNIT 1 3/4 6-26 dUL 23 1975

. . [ -

[ CONTAltiMEilT SYSTEMS SHIELD BUILDI!!G STRUCTURAL INTEGRITY LDtITING CO.'IDITIO!! FOR OPERATIO:t 3.6.6.3 The structural integrity of the shield building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.X.3.

g .

APPLICABILITY: MCUES 1, 2, 3 and 4. a ACTION: ,

hith the strt:ctural integrity of'the shield building not conforming to 1

. - j the original acceptance standards, restore the structural integrity to 1 within the limits prior to increasing the Reactor Ccolant' System temperature above 200'F. ,

. . . l m

SURVEILLA:lCE REOUIREMEllTS 4.6.6.3 The structural integrity of the shield building shall be deter- .

mined during the shutdown for each Type A containment leakage rate test I (reference Specification 4.6.1.2) by a visual inspection of the accessible j interior and exterior surfaces of the shield building and verifying no  :

apparent changes in appearance 'of the concrete surfaces or other abnormal

. degradation. Any abnormal degradation of the shield building detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.1.

t DAVIS-BESSE, UNIT 1 3/4 6-27 dli. 2 ? e -

4 .

TABLE 4.6-1 ,

ACCESS CPE?tI!;GS REGUIRED TO BE CLOSED TO EtiSURE SHIELD EUILDING INTEGRITY I. AIR TIGHT COORS DOOR t;0. DESCR!PTICN ELEVATIO!1

~

100 Access Door frem the No.1 ECCS Pump Room 545' (RcomjlC5)toPipeTunnel101

. 104A Access Coor frem Stair AS-3 to the No.1 555' ECCS Pump Recm (Recm 105)

. 105 Access Dcor frem Passage 110A to the area above 555'

,the, Decay Heat Coolers .

107 Access Door frca the No. 2 ECCS Pump Room 555'

, (Room 115) to the Miscellaneous Waste Monitor .

Tank and Pump Room (Reca 114) 108 Access Door from the No. 2 ECCS Pump Room 555' (Room 115) to the Det;ergent Waste Drain Tank .

and Pump Rocm (Rcom 125) 201-A Access Door from Corridor 209 to the No. 1 565' Mechanical Penetration Room (Room 208) 204 Access Door from Passage 227 to the Makeup 565' Pump Rcca (Rcom 225) 205 Access Dcor frca Passage 227 to the No. 2 585'

. Mechanical Penetration Room (Room 236) 308 Access Ccor from Corridor 304 to the No. 4 585' Mechanical Penetration Room (Room 314)

II. BLOWOUT PAllELS .

TOTAL NO. ' LOCATION ELEVATION 1 . No. 2. Mechanical Penetration Room 565' (Roca 236) 6 No. 3 Mechanical Penetration Room 585' (Room 303) 6 No. 4 Mechanical Penetration Room 585' (Roca 314)

DAVIS-BESSE, UNIT 1 3/4 6-28 s c i f 373 4 1

1 PLAtlT SYSTEMS

; MAlti STEAM LIl:E ISCLATIC1 VALVES

• LIMITI!:G C07:01 TIC ; FCR CFERATICi!

3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MCCES 1, 2 and 3.

ACTIO!:

~

MODE 1 - With cce r.in steam line isolation valve inoperable, POWER OPERATIO:t ray continue provided the inoperable valve is either -

restored to 0FERASLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

.  : Otherwise, be in HOT SHUTCCW:1 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

t MODES 2

. and 3 - With one main steam line isolation valve incperable, subsequent operation in MODES 1, 2 or 3 may proceed provided: .

a. The inoperable isolation valve is maintained closed.'

(

i Otherwise, be in HOT SHUTCOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

^

SURVEILLA!!CE REGUIRE:'E::TS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLEbyverifyingfullclosurewithin$secondswhentestedpursuant to Specification 4.0.5.

7K .

l t.

DAVIS-BESSE, UtilT 1 3/4 7-9 l

o-1

TH:S ?AT 714

' 3 'EE U OF id:F 0 :!;.' ... . . . . .. . . i '.U N I 1

PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ,

LIMITIG CONDITION FOR OPERATIO*1 3.7.7.1 Two independent control recm emergency ventilation systems -

shall be OPERABLE. .

APPLICABILITY: MODES 1, 2, 3 and 4.

ACT'ON:

With one control room emergency ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'.iN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -

SURVEILLANCE REOUIREMENTS 4.7.7.1 Each control room emergency ventilation system shall be demon-strated OPERABLE:

I I

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room c[od22I air temperature is 1,(120)*F. I

b. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes and

'2. Verifying that each ventilation systen is aligned to receive electrical pcwer frem, separate OPERABLE essantial busses. .

c. At least once per 18 renths or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system, by:

JUL 0 0 076 DAVIS-BESSE, UNIT 1 3/4 7-17 e

REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - AXIAL POWER SHAPING R03 GROUP LIMITING CONDITION FOR OPERATIO'l 3.1.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless -

fully withdrawn, and shall be positioned within + 6.5% (indicated position) of their group average height.

APPLICABILITY: MODES 1* and 2*. s ACTION: .

With a maximum of one APSR inoperable or misaligned from its group average height by more than + 6.55 (indicated position), operation may continue provided that within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

a. The APSR group is positioned s'uch that the misaligned rod is restored to within limits for.the group average height, or ,

b. It is determined that the imbalance limits of Specification 3.2.1 are satisfied and movement of the APSR group is prc-vented while the rod remains inoperable or misaligned.

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 positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Asyr=etric L.~ v1 Rod '::..'.r.- is inoperable, then verify the v*mee-5 g5dc2 .m; ;;;.M :ms at le t once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (, een w,7 3 nz Foutt- %vetry metcpick 4.1.3.2.2 Unless all APSR are fully withdrawn, each APSR shall be rever c,acygt.

determined to be OPERABLE by moving the individual rod at least 2% ,o 33 vgm4 at least once every 31 days. rSe a.uwir

• See Special Test Exceptions 3.10.1 and 3.10.2.

I' .

s.

xs ,

DAVIS-BESSE, UNIT 1 f 3/4 1-21 -

i

,, REACTIVITY CONTROL SYSTEMS ,

POSITION INDICATOR CHANNELS LIMITING CONJITION FOR CPERATION-1 3.1.3.3 All safety, regulating and axial power shaping control ttd absolute position indicator channels and relative position indicator channels shall be OPERABLE and capable of determining tne contrei rod positions within + 6.5"

~ '

APPLICABILITY: MODES 1 and 2. .

ACTION:

With a maximum of one absolute position indicator channel per control

~

rod group or one relative position indica'tcr char.r.el per control rod group inoperable, witnin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reduce TAERMAL PC'.,ER to < 60S cf the THERMAL POWER allcwable for the reactor coolant pump ccmbination ens reduce the Nuclear Overpower Trip Setpoint to < 70L of the THER.".AL -

POWER allowable for the reactor coolant pump com,bination.

SURVEILLANCE RE0'JIREMENTS 4.1.3.3 Each absolute and relative position indicator chcr.nei -hall be determined to be CPERABLE by verifying that the absoluta positica indicator channels and the relative positica indic: tor channels agree with 6.55 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time inter.cls when the Asymmetric Rod 'b ';sr is inoperable, then ccmpare the auseiute position indicator an relative position indicator channels at le;;t once per 4 hcurs.

Fav W cwcu@y

^

DAVIS-SESSE, U.11T 1 3/4 1-22 ,, . , ,

.* a.:

I l

REACTIVITY CONTROL SYSTEMS ,

R0D DROP TfME ,

LIMITING CONDITION FOR OPERATION __

8 3.1.3.4 Th'e individual safety and regulating rod drop time from th?

fully withdrawn position shall be < l.65 seconds from power in:o ruption at the control rod drive breakers t'o 3/4 insertion (25q position) with:

. a. T,yg a 525'F, and Mk2 ,

b. All reactor coolant pumps operating. .

A'PPLICABILITY: MODES 1 and 2.

'CTION:

a. Withthedroptimeofanysafetiorregulatingrodcatermined to exceed the above limit, restore the rod drop tire to within the above limit prior to proceedi.ng to MODE 1 and 2.

b. With the rod drop times within limits but determined witn less than 4 reactor coolant pumps operating, operation ray src:aed provided.that THERMAL POWER is restricted to less tnan 3r equal to the THERMAL POWER allowable for the reactor e.e,ciant

, pump ccabination operating at the time of rod drop ti,te l measurement.

1 l

SURVEILLANCE REOUIREMENTS 1

I 4.1.3.4 The rod drop time of safety and regulating rods shall be demon- i strated through measurement prior to reactor criticality:

1

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

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

c. At least once every 18 months. l DAVIS-BESSE, UNIT 1 3/4 1-23 j t' '

s. \

4 l

I 5I

RECTI'.'ITY CC'.7 L SYSTE'*3 5; n 7.c.. y;----- _- -  % -

LI"ITI' 3 ::':3 T!C': F0, OSE ATIO::

3.1.3.5 AI! safety rcds shall be fully withdra::n.

A??LI C'J :'_ ITY: 1* and 2*#,

ACTIO'::

With a nr .

~

s DAVIS-CESSE, U"IT 1 3/4 1-24

~

JUL t. t k:. .

I i,

9 f

REACTIVITY CCNTROL  ; JW i ; LI"IIS SYSTEMS f!  !

m-TEGULAT!"3 ROD ' -

11"ITI:;G C0"O*TIO'l FOR CPERATIO'!

, r itsu in

- r gulating rod groups shall be 3.1-3 s 3.1-1 and Jim with7 a rod group 31'6 The te5irtic" as she;;n i l withdrav.n grcups 5 and 6/ .

pi;ysical on Fig;re overlap of 25 + 55 between sec.uent a MODES 1* and 2*!. .

s APPLICABILITY:

ACTIC'!:

beyond the above,insertien limits r regulating red groups inserted cutsrication 1;dethespecifiedlimits,excee

• .1.3.1.2, eit er.

1Uith the

'or with any group sequence i hin the erlimits overl:pfor within'2 sur

~

Restore the regulating groups to w t

a. l to that fraction,of hours, or Reduce THEP..AL P0h'ER hours, or to less than or equa

b. RATED THEP.:4AL POgER , which is allowed by using the above figurcs withinhours. 2

c. Be 'in a't least HOT STAT:DBY within 6 1 and 3.10.2.

• See Special icst Exceptions 3.10.

,fwith Keff > 1.0.

t.'J; - .  :

3/4 1-25 II'""'c crRSE. U"IT 1 N!

F REACTI'!ITY COCOL SYSTElG f

REG'JL AT!'.G ROD O.~ ~"R, L if t:TS W WW

m. ,

l 1

, SUTNEIlL A*:CE RE0;RE"E."TS 4.1.3.6 The p:sjtiIo/$$wu~ad f er:h resultting group sh:ll be detemined to be within the in::r.ica, sequence and overlap liraits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except when:

a. The regulating red insertion limit alarm is inoperable, ti;en verify the grcups to be within tha insertion lir::its at least ,

once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;

b. The control rod drive sequence alarn is inoperable, the1 .

Verify the groups to be within the sequence and overlap limits at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. '

D O

e 9

9 t

e e

4 9

9 e

DAYlS-BESSE, UNIT 1 3/4 1-26

, s e.

t i

i i

e

-i THis p/, ,tg 7; n: -

a .cu ._,5tPT OF Ip'e y . -

f .s  :::

0 /dFUC'.NT

, .... . . A # T"_ ' gyg I

• I .

1 i

i 4

1 l 1-J Figure 3.1-2 REGULATING Rod Groun t!ithdrawal Limits for 4 Pump Operation up to Control Rod Interchange (250 + 10 EFPD)

DAVIS-BESSE, UNIT 1 -

3/4 1.-27 DUN 25 g e

--,,w--e---n ,.,,----.-w,,-e ,-- --,-g-, -- ,em,,---.-- eeme e,-,--

i

, 7-,.

s j

t -b

. '.:,. o..c .P.o. T O F

- 8 '

THis P/.. -

PUCANT ggpo r....  : 0: 4,;i ns-

. r,mn-

.. . nC -

1

(

i 1

1 i

( -

l . ,

. Figure 3.1-4 .,

l Control Rod Core Location and Group Assigre.ents up to (2C: ; ~.?) EFPD

?Co s to -

DAVIS-BESSE, U'iIT 1 3/4 1-30 OUI 2 51975

• a

'I 4

,n. .

k

=

4 *

( .

i .r . . T. OF

...-. .ra -

T Hi "~ P .-r,. ,. . . ,. . -..

c/dPUCANT tNO ?.:n.- n

'N.-.- 'YT$UN -

t .

1-I -

i Figure 3.1-5 t

Control Rod Core Location and Group Assignments after (?? i!?) EFPD .

200 40 s

DAVIS-BESSE, UtilT 1 . 3/4 1-31 JW;:: .

3/4.2 POWER DISTRIEUTION LIMITS AXIAL POWER IMEALANCE

<~

s- .

LIMITING CONDITION FOR OPERATION 3.2.1 on AXIAL3.2-1 Figures POWERand IMBALANCE 3.2-2. shall be maintained within the limits shown APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.*

ACTION: "

With AXIAL POWER IMBALANCE exceeding the limits specified above, either:

a.

Restore the AXIAL 15 minutes, or POUER IMBALANCE to within its limits within

b. Be in at least HOT STANDBY within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

k '

SURVEILLANCE REC'JIREMENTS M

4.2.1 The AXIAL POWER IMEALANCE shall be determined to be wi.hin limits in each core quadrant at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when abov. 403 of RATED THERMAL POWER except when an AXIAL POWER IMSALANCE ::.'# t:r is inoperable, then calculate the AXIAL POWER IMBALANCE;i" ::ch :;r; quadr:nt "ith ea inep:r;ble ..~.4;or at least once per hour. *

= See Special Test Exception 3,10,1 DAVIS-BESSE, UNIT 1 3/4 2-1 JUL s ,.

O ' '

\.

( ,

TH'i FA c .? 'n :2..: .; . ; -aEcs:pr o,e INFOKin f.Ot4 FROta. InE APPLICANT aloT , Puc.

l 1

4 AXIAL POWER IMSALANCE ENVELOPE .

, FOR OPERATIO:4 UP TO (400 = 10) EFPD Figure 3.2.1

. DAVIS-BESSE, UNIT 1 3/4 2-2 JUL 0 91976

+

.ne r

O

=> e, O ,

s

\ -

7HG F'.O G?EN P5N: : 7. ,', -',.. .nI ,r ..,w i

INTO".c.'..,T:ON FROM i;,2 A, pt!cApg A>c r reog _

I t

l AXIAL PCWER IMBALAMCE ENVELOPE l FOR CPEPATICi AFTER t400 =.10) EFPD l i Figure 3.2-2 DAVIS-EESSE, UNIT 1 3/4 2-3 Jut, 0 3 ;3., j e

v - -, _ , . , - _ ,_, , . . . . , _ , _ - _ _ _ . _ _ . . . I!

i P0'n'ER DISTRIBUTION LIMITS, em SURVEILLANCE REGUIREMENTS (Continued)

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.

c. The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 The measured F of 4.2.2.1 above, shall be increated by ,34F I*

to account for manufact0 ring tolerances and further increased by.)y-to account for measurement uncertainty.

7, cj[

f I

l l

l i

I l

DAVIS-BESSE, UNIT 1 -3/4 2-5 JUL 0 01976 l'

POWER DISTRIEUTION LIMITS rR SURVEILLANCE REQUIREMENTS 4.2.3.1 FN shall be determined to be within its limit by using the incore det$ctors to obtain a power distribution map:

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

b. At least once per 31 Effective full Power Days. _

c. The provisions of Specification 4.0.4 are not applicable. ,

S Yo N

4.2.3.2 The measured F of4.2.3.1above,shallbeincreasedby({f) formeasurementuncertaiNty.

e i .

e 4

6 DAVIS-BESSE,0 NIT 1 3/4 2-7 Jul. 0 3 u,q.

e e

i l

l l

POWER DISTRIEUTIO;l LIMITS QUADRANT POWER TILT LIMITING CONDITION FOR OPEPATIO!!

3.2.4 THE QUADPA!!T POWER TILT shall not exceed [.

APPLICABILITY: MODE 1 above 15% of RATED THERMAL POWER.*

ACTION:

% gh ,

a. With QUADPANT POWER TILT determined to exceed 4 (_4) but < d

1. Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

a) Either reduce the QUADRANT POWER TILT to within

- its limit, or b) Reduce THERMAL POWER so as not_to exceed THERMAL

. POWER, including power level cutoff, allowable for

, the reactor coolant pump combination less at least

\

2% for each % of i i = -i QUADRANT POWER TILT in y'ga_ excess o and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the High t-tux Trip Setpoint and the Flux - a Flux - Flow

' Trip Setpoint at least 2% for eac 1% of indicated QUADRANT POWER TILT in excess of 4%' %

2. Verify that the QUADRANT POWER TILT is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER to less than E05 of THERMAL POWER allowable -

for the reactor ccolant pump ccmbination within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the High Flux Trip Setpoint to <

. 55% of THERMAL POWER allowable for the reactor coot-ant pump combination within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,

3. Identify and correct the cause of the out of limit con-dition prior to increasirg THEPSAL POWER; subsequent POWER OPERATION above i..; of THERMAL POWER allowable for the reactor coolant pump cc:bination may proceed provided that the QUADPANT POWER TILT is verified within its limit at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-8 -

JdL 0 0 36

POWER DISTRIEUTIC!i LIMITS SURVEILLA' CE RECUIREME!!TS 4.2.4 The C'JACFA T FC'aER TILT shall be determined to be within the limits at least once every 7 days during operation above 15% of RATED THERMAL POWER exceot when the QUADPANT POWER TILT n= tr is inoperable, then the QUADR;JiT FC*aER TILT shali be calculated at 'dast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

( -

DAVIS-BESSE, UNIT 1 3/4 2-10 JU! ..

i l

L. t

i TABLE 3.3-1 (Continued)

/* -

TABLE NOTATION

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

• • When Shutdown Bypass is actuated.

1. The provisions of Specification 3.0.4 are not applicable.

IfHigh voltage to detector may be de-energi::ed above 10-10 amps on both Intermediate Range channels.

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

4 (1) The Nuclear Overpower Trip Setpoint is 1 5% of RATED THEPRAL

- POWER, (2) The Shutdowr. Bypass RCS Pressure--High Trip Setpoint of < 1820 psig is imposed, and y

. (3) The Shutdown Bypass is removed when RCS pressure > psig.

ACTION STATEMENTS

' ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel' to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDSY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the control rod drive trip breakers.

ACTION 2 -

With the number of OPEPABLE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION may proceed provided all of the following conditions are satisfied: ,

a. The inoperable channel is placed in the tripped

. condition within one hour.

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, O

DAVIS-BESSE, UNIT 1 3/4 3-3 JUL g ,. ,

l

- I o

TABLE 4.3-1 -

. REACTOR PROTECTI0tl SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

.e E

" 1 Y - . CHAtlNEL MODES IN WHICH E CHAlltiEL CHAllNEL FurtCTI0t1AL SURVEILLAttCE FUNCTI0flAL UllIT CilECL CALIBRATI0il TEST REQUIRED

~

Manual Reactor Trip N.A. N.A. S/U(1) N.A.

g 1.

Z 2. High Flux S D(2),andQ(7) M 1, 2

3. RC liigh Temperature S' R H 1, 2

4. Flux - AFlux - Flow S(4) M(3)andQ(7,8) M 1, 2

5. RC Low Pressure S R M 1, 2

6. RC High Pressure S R M 1, 2

7. RC Pressure-Temperature S R M 1, 2

8. liigh Flux /flumber of Reactor 1, 2

$ Coolant Pumps On S R M 1, 2

'y 9. Containment liigh Pressure S R M

10. Intermediate Range, Neutron -

Flux and Rate S R(7) S/U(5) 1, 2'and*

". Source Range, Neutron Flux .

and Rate S' R(7) S/U(1)(5) 2, 3, 4 and 5

"-^1-Rod-Dr4ve=TMir= Breakers +!-fr. -th4h t and L/3ft). - 1, -0 =P 13.' Reactor Trip Module Logic H.A. N.A. M 1, 2 and*

14. Shutdown Bypass High Pressure S R H 3, 4 and 5**

[=< .

• 4 S=m J' -

I TABLE 4.3-1 (Continued) -

NOTATID';

• - With control r:d drive trip breaker closed.

l l

• • - When Shutdown Bypass is actuated.

(1) - If not perfor ed in previous 7 days.

(2) - Heat balance only, ab:ve 15% of PATED THER':AL F J..

• pgawcNdL TW .shatt b e a,ervsect f<cm nhos CALISW'0" '

. (3) - Compare ir.: Ora to excore reasured AXIAL PC;lER IMBALA! ICE above of RATED THER*'.AL POWER. Recalibrate if abs.olute difference "

ro %(M

> percer.t. .%. MJ 'M M 4<

.T. s 6 h Job AXIAL PO'.,'ER IM3ALA!!CE and 1006 flow indications only.

d a' d J ~:

(4) -

. (5) - Verify at least ene decade overlap if not verified in previous 7 days. .

(6) - Each train tested every othen enth.

(~ Neutron detectors may be excluded froar CHAf!!!EL CALIBRATI0ff.

(7) -

(8) -

Flow rate reasurement sensors may be. excluded from CHA!!!1EL CALIBRATICi;. However each sensor shall be calibrated at least once ,per 18 c:ntns.

/

THIS PA 'E 1:EN .s: i '3 RECElPT OF INFO 9;.'. l ? i-P . e '. 1 E A. PLICANT DAVIS-BESSE, U!iIT 1 3/4 3-8 JUL 0.' i.-..2 e

l

. l

~

TABLE 3.3-3'

. SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION e

32 Lee MIfilMUM Looc i T4 TOTAL fiO. CHAfitlELS CilAfitiELS APPLICAB!E -

6 FUNCTIONAL. UNIT OF.CilAfif1ELS TO TRIP OPERABLE  !!0 DES ACTION k 1. SAFETY IllJECTION

'#6'C c- a. liigh Pressure Injection .

2 Z 1)ContainmentPressure-High ,4 2 3 1,2,3 ,84

2) RCS Pressure - Low 4 ,2 3 1,2,3 9Y

b. Low Pressure Injection

' 1) Containment Pressure - High 4 2 3 1,2,3 g1

, 2) RCS Pressure--Low-Low ** 4 -

2 3 1,2,3 A7 Rs .

w 2. CONTAINMENT SPRAY

$ a. Containment Pressure--High-High 4 2 3 1,2,3 E' <y

3. CDNTAlf1 MENT ISOLATION .

a. Containnient Pressure - High .

4 . 2 3 1,2,3 4f

b. Containment Pressure--High-High 4 2 3 1,2.3 Sy

, c. Containment Radioactivity - High 4 2 3 All. MODES E and

• D/0

d. RCS Pressure - Low

• 4 2 3 1,2,3 3?

,5 o .

'l

' s

/

~ ' '

TABLE 3.3-3 (Continued)

. SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

?

dO6ic- MItoIMUM dosse g TOTAL NO. CilANflELS CilAfiNELS APPLICABLE g FUNCTIONAL UNIT OF _CilAllflELS TO TRIP OPERABLE MODES ACTION

3. CONTAINMErlTISOLATION(continued) dG/c

e. RCS Pressure--Low-Low ** 4 2 3 1,2,3 59 ,

4. CONTAINMENT COOLING

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

b. Containment Pressure--High-High 4 2 3 1,2,3 S9 m c. RCS Pressure-Low 4 2 3 1,2,3 p7 1 .

Y, 5. MAIN STEAM AtlD FEEDWATER ISOLATION .

~

a. Containment Pressure--High-High 4 2 3 1,2,3 Ji ?

6. CONTAINMENT SUMP SUCTION -

a. Borated Water Storage Tank - Low 4' ,

2 '3 1,2,3 8y C

3-

"im

~

e eme-O

• n>

. i 0

TABLE 3.3-3 (Continued) ,

SAFETY FEATURES ACTUATION SYSTEM INSTRIMENTATION g

• 8 U MIrlIMUM En TOTAL NO. CilANNELS CllANf4ELS APPLICABLE OF CHAtit1ELS TO TRIP OPEPAULE MODES ACTION

@ FUNCTIONAL UNIT m

~

' E

7. SFAS MANUAL ACTUATION -

CHANNEL A 1 1 1 1, 2, 3, 4 W" '

Q

8. SFAS MAtlVAL ACTUATION -

M 81 CilANNEL B l 'l 1 1, 2, 3, 4

9. CONTAltiMEtiT SPRAY MANUAL ACTUATION CilANNEL 1 1 1 1, 1, 2, 3 Fr il R

10. CONTAlt4 MENT SPRAY MANUAL ACTUATION CIIANilEL 2 l 1 1 1,2,3 16 li

[

h

11. COINCIDENCE LOGIC CHANNELS 34 . 2*** 2*** All Modes M /J

12. SEQUE!4CE LOGIC CilATHIELS 4

2 3 1,2,3 M9 4

k,- -

l .

I TABLE 3.3-3 (Continued)

TABLE fiOTATIO!! ,

• Trip function may be bypassed in this MODE with RCS pressure below 1800 psi . By; ass shall be autcmatically removed when RCS pressure exceeds 1E00 psig.

• • Trip fur.cticn ray be bypassed in this MODE with RCS pressure belcw 600 psig. Eypass shall be automatically removed when RCS pressure exceeds 600 psig.

ACTION STATEMErlTS

=

A ACTIO::g- With the number of OPERAELE channels one less than the -

Total Number of Channels and with RC system pressure.

a. ' < 1800 psig, place the inoperable channel in the bypassed condition within one hour and restore the inoperable channel to OPERABLE status within 24

- hours after increasing the RCS pressure above 1500

. . psig; otherwise be in a.t least HOT STANDBY.within -

- g' the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

( . b. > 1800 psig, place the inoperable channel in the Fypassed condition within one hour; operation nay

. continue provided that the Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4. hours for surveillance-

- testing per Specification 4.3.1.2.

With less than the Minimum Channels OPERABLE, operation ACTION )( -

~

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

8 ACTION 1)Q - With less than the Minimum Channels OPERABLE, be in at least HOT STAND 3Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. /

l>

ACTION N - With any comporient in the coincidence logic channels-inoperable, trip that component within one hour or be in at least HOT STANC3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'.'N within the follcwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l

( .

DAVIS-BESSE, UNIT 1 3/4 3-13 l l

H

a t

,~,

a e

e I.

It *t-

'. t e s-c.,;,.e QC t s,'

.e. .* w .4..

. -s

r. vs , g y,-g ,, g,.-

iC pg. r LI .. *'* *. I[

, Mn'r > P OE Figure 2.1-2 Reactor Core Safety Limit DAVIS-BESSE, UNIT 1 2-3 l

1 i i .

e a

g i

TABt.c 2.2-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS '

E 5 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES T

R 1. Manual Reactor Trip Not Applicable flot Applicable '

.m 2. High Flux < 105.44% of RATED THERMAL POWER < 105.5" of RATED THERMAL POWER !

g Uith four pumps operating Uith four pumps operating 7

U 5g7 gp 1,80.7% of RATED THERMAL POWER <(80.7:hofRATEDTHERMALPOWERf As p [s b.

m with three pumps operating ,

with three pumps operating j Sr . /g < 53.0% of RATED THEP. MAL POWER with

< (53.0dof RATED THERMAL POWE i one pump operating in each loop one pump operating in each loop i *

3. RC High Temperature < 618.95 F < 619*F

4. Flux - a Flux-Flow (I) Trip Setpoint adjusted to not Allowable Values not to exceed i m exceed the limit line of the limit line of Figure 2.2-2. '

& Figure 2.2-1. l S. RC Low Pressure (I) 1 985.4 1 psig 1 1985 psig .

6. RC P e sure U) k 2356 psih A3 O ' 0 '

< 2355 psig *

7. RCPressure-Temperature (I} > (13.85 Tout F - 6494) psig > (13.85 Tout F - 6498) psig THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM ThG APPLICANT

• O e

is.

I l

g. -

. . e ., n.n.nr - . , , g. . . '-

g n y: . 3, . ,

z; . . , . . - , : '.. ; ..'. ; .'

.. .. r w . ..

. . - - ~ -

-->;,c ...y,3=,

_d. N

. :4 fi 7' T

...e 9 -  : La y

,t3i.l.?.3 k.

_ TABLE 2.2-1 E

3 REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS -

'f FUNCTIONAL UNIT TRIP SETPOINT M ALLOWABLE VALUES.

M m

1. Manual Reactor Trip '

Not Applicable Not Applicable

2. liigh Flux (I) 1 1 05.44% of RATED THERMAL POWER

~

5 105.5% of RATED THERMAL POWER

3. RC High Temperature 'I < 618.95*F *

< 619*F

4. Flux- A Flux - FlowI2) . Trip Setpoint adjusted to not exceed the limit lines of Figure 2.2-1 Allowable Values not to exceed '

- the limit lines of Figure 2.2-2.

l '? W

' " When the liigh Flux trip setpoint is required to be reduced by an ACTION statement j to some percentage of the THERMAL POWER allowable for t..e reactor coolant pump combination, the TilERMAL POWER allowable:

a. For 3 pump operation, is 80.7% of RATED THERMAL POWER.

. b. For operation with one pump in each loop is 53% of RATED THERMAL POWER.

(2)that:Trip may be manually bypassed when RCS 1 pressure 1 820 psig by actuating Shutdown B "

,. f gg a.. The thie+ car-hc;; .;c;- trip set point is 1 5% of RATED THERMAL POWER.

b. The Shutdown Bypass High Pressure trip is automatically imposed with a set. ;oint -

< 1820 psig, and

c. ' The Shutdown Bypass is manually removed when RCS pressure >,.MOIT.

~p 19 R S. 4 o

N

~* -

. 3, - . _ . . _ . _ - - - - - -

t.

i

/

1 lI e

I- THR' - ' .3 9E E.'PT OF

' I N r'.'.s

... . . . ..I c A.- FUCANT NOT rRoc l 1

i I

Figure 2.2-1 I

,.,____._,__......,_,_n.m.--._ o , e s -,, e re e, ,,.

l, sasFdr"' +'-r..,_u',~.----

o..u m. um r v..-a . . .. . . s -. 2 0 -

Flut .a Flux FLDu) 4.me .StrpaturS I

DAVIS-BESSE, UNIT 1 2-7 .

JUN 9 41978 1

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

l.

d

n i .

a t

i N

.l 4

i THis FA . - -

. .: ?T O F IN FO RM... r,.-);., 3 ., .

,gog_

unr I .

Figure 2.2-2

. . , , _ _ _t , - . . _ , , . , _ , - _ , _ _- --__. - ....

. n..unau.m ,o.um . . .. .um- ..--..r.,.....--

v . vi ..ui usaum v.. nuo

, .ma ci.u , u . 2, su ,- v .. ua i . .a n n..uc-  ;

flux -4nn - Row The Allowohk Mes DAVIS-BESSE, UNIT 1 2-8 JU?! 24 ;976

i 2.1 SAFETY LIMITS m .

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. Overheating of the fuel cladding is prevented by rest.-icting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and "

the cladding surface terperature is slightly above the coolant saturation temperature. .

Operation above t!)e upper boundary of t'he nucleate boiling regime

. would result in excessive cladding temperatures because of the onset of departure frca nucleate boiling (DN3) and the resultant sharp reduction in heat transfer coefficient. D;B is not 9. dire:tly measurable parameter during operation and therefore ThEoMAL POUER and Reactor Coolant Temper-c ature and Pressure have been relat ad to D:!B through the B&W-2 DNB correlation. The DNB correlation has been developed to predict the DNB axially uniform and non-uniform heat flux and the location of DNS .~m flux distributions. The 1~'a'. i NB heat flux rhtio, DNBR, defined as the i wuld cause DNB at a p' articular core location

. ratio of the heat flux t" l to the local heat flux, i indicative af the margin to DNB.

I The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.32.

This value corresponds to a 95 percent probability at a 99 percent

! confidence level that D.'IB will not occur and is chosen as an appropriate

- margin to DNB for all operating conditions.

The curve presenteu in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.32 is predicted for the maximum possible thermal pcwer i 1127, when the reactor coolant flow is 131.3 x 106 lbs/hr, which is the

! design flow rate for four operating reactor coolant pumps. This curve is

! based on the following nuclear power peaking factors with potential fuel i

densification effects:

\ '

= 2.56; Fh=1.71; F = 1.60 i S #j The design limit power pe Sing factors are the most restrictive calculated at full powerj *'r the range from all control rods fully withdrawn to minimum allo. ble control rod withdrawal, and form the a

core DNBR design basis.

., j \.

l .-

l j DAVIS-BESSE, UNIT 1 B 2-1 JUN 2 4 jfg ,

{

i .~ .t:

t .

l

SAFETY 1.IMITS r

BASES

- The reactor trip envelope appears to apor ach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about ( ) psi less

• core outlet pressure, providi; i more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based cn the more restrictive of two themal limits and include the effects of potential fuel densification: 4

1. The 1.32 0:iBR limit pr,oduced by a nuclear power peaking factor of k = 2% or the combination of the radial peak, F: axial peak and position of the axial peak that yields no ! css 4 Z.56 than a 1.32 D:iSR. , <

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.4 kw/f t.-

Power peaking is not a directly observablp 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 correspond to the expected minidun ficw rates wiUi four pumps, ti.ree pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor cootnt pump-maximum thernal power combinations shown in BASES

• Figure 2.'. The curves of BASES Figure 2.1 represent the conditions at

! uhich a r..;nimum D:;SR of 1.32 is predicted at the maximum possible themal pctier for the number of reactor coolant pumps in operation or the local qualicy at the point of minimum D:iSR is equal to +22%, whichever r

condition is nore restrictive. .

Using c local quality limit of +22% at the point of minimum D:iBR

as a basis for cerve 3 of SASES Figure 2.1 is a conser ".tive criterion

even though the quality at the exit is higher than the luality at the point of ninimum D

iBR.

t

The D::BR as calculated by the B&W-2 D:iB correlation continually increases from point of mininum D:;BR, so that the exit D:;3R is wtays

. higher. Extrapolation of the correlation beyond its published quality

range of +22% is justified on the basis of experimental data, i .

t i DAVIS-SESSE. U:llT 1 B 2-2 4L'i 2 4 g n .

\

,1 i

I i

TH:5 P '

SAFETY LIMITS ,, .qg, g;p7 . cyp f ,.

IN, m, ., ,*

\.

i

• u su FL: f.NT BASES For each curve of BASES Figure 2.1, a pressure-ter?erature point above and to the left of the curve would result in a DNSR greater than 1.32 or a local quality at the point of minimum C"3R less than +223 .

for that particular reactor coolant pump situation. The 1.32 0:lSR curve for four pump operation is core restrictive than any other reactor J ccolant pump situation because any pressure / temperature point above and to the lef t of the four pump curve will be above and to the lef t  :

of the other curves.

^

2.1.2 REACTOR CCOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Ccolant System frca overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment etmosphere. .

- The reactor pres.ure vessel and pressurizer are designed to Section i*

III of the ASME Boiler and Pressure Vessel Code which percits a maximum transient pressure of 1103, 2750 psig, of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Ici62 Edition, which permits a maximum transient pressure of 1100, 2750

+ psig, of cc:ponent design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psig,125" of design pressure, to demonstrate integrity prior to initial operation.

1 i

4 1 .

't s . .

V.

i ,..

l DAVIS-BESSE, UNIT 1 -

B 2-3 .

.! '. JUN g 4 .

1975 ..

I -

)

TM!3 :'

~

~:! N I "'I O E I'II C'}

l p , ., . .

. . , ; . .. l . .E h?P'EAl I f 2.2 LIMITII:G SAFETY SYSTEM 2E" ** "

( n.

BASES 2.2.1 REACTOR PROTECTIO:i SYSTE!! !::STF.UME!;TATIO:1 SETFOI:lTS .

The Reactor Protecticn Systc.- Instrumentation Trip Setpoint specified ,

in Table 2.2-1 are the values at Aich the Reactor Trips are set for each parameter. The Trip Setpcints have been selected to ensure that the reactor core and reactur ccolar.t system are prevented from exceeding their safety limits. Operatice, with a trip setpoint less conservative  ;

than its Trip Setpoint tut withir. its specified Allowable Value is accept-able on the basis that each Allowable Value is equal to or less than the .

drift allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certcin functions of the Reactor Protection Systc . in order to permit centrol rod drive tests, zero pc.ter FHYSICS TESTS art certain startup and shutdcwn procedures.

The purpose of the Shutdcun Bypass High Pressure trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal icw pres.,Jre trip setpoint so that the reactor cust be tripped before the bypass is initiated. The High Flux Trip Setpoint of < 5.0; prevents any sigriificant reactor power from being produced. Yufficient natural circulation would be available to remove S.0!' of RATED THEFRAL F0WER if none of the reactor coolant pumps were operating. -

Manual Reactor Trip .

The Manual Reactor Trip is a redundant channel to the automatic j Reactor Protection System instrumentation channels and provides manual i reactor trip capability.

l l

• High Flux i A High Flux trip at high power level (neutron flux) provides reactor core protection against . reactivity excursions which are too rapid tio be protected by temperature and pressure protective circui "y .

' During nomal station operation, reactor trip is initiate' when the reactor power level reaches 10S.S5 of rated pcuer. Due to calibration and instru-ent errors, the caxicum actual power at which a trip would be actuatedcouldbe)l12)%,whichwasusedinthesafetyanalysis.

~ ~

l r

DAVIS-SESSE, Ulli 1 B 2-4 JUy g4

1975

~

~

l

. i 1.!:i!T**:3 SUETY SYSTE" SETT!':3S

^ .

V BASES RC Hi::5 Tencerature The :tC Pigh Temoerature trip e 619'F prevents the reactor outlet te.,? erasure frc, exceeding the dest;n limits ar.d acts as a backup trip for all p:. tar excursion transients.

. 4

• ~

Flex - 1 Flux-Flow .

The power level trip setpoint produced by the reactor coolant system flow is based on a flax-to-flow ratio which has been established to accc::cdate flow decreasing transients froa high power where pro-tection is not provided by the High Flux / Number of Reactor Coolant Pumps On Trips.

The power level trip setooint produced by the power-to-flow' ratio providcs both high poaer level and low flow protection in the event the

( reacter pcaer level increases or the reactor cco.lant flow rate decreases.

The power level setpoint produced by the power-to-flow ratio provides overpcwer CH3 protection for all tedes of pump operation. For every ficw rate there is a maximum permissible poaer level, and for every power level there is a minimum permissible icw flow rate. Typical power level and icw flew rate cc Dinations for the pump situations of Table 2.2-1 are as fo11cws: g g

1. Trip uculd o ar when fourkeletor coolant pumps are operating if power is 108.05 a re' actor ' low rate is 1005, or flow rate is 92.65 and cwe le el is 1005.

2. Trip would o ur wheg threb reactor c lant pumps are operating if pcwer ise 0.75 gnd reactor flow rate 1 '4.7%, or flow rate is 69.45 and power isg75::. -

3. Trip wculd cccur when one reactor ccolant pump is ope ting in each icep (tota' of two pumps operatir.g) if the power isj52.9%

and reactor fl a rate is '9.0?; or ficw rate is 45.4% and the powerlevelisj49.05.

For safety calculations the caxicum calibration and instrumentation errors for the pcwer level were used.

l DA'!IS-SESSE, UNIT 1 B 2-5 JUli 2 8103 i

LIMITit G SAFETY SYSTEM SETT!*:35 r

i BASES" The AXIAL PC'.ER I:iEALA*:CE t0undaries are established in order to prevent reactor thermal lini:s fecn being cxceeded. These taermal limits are either power peaking kw/f t limits or D::3R limits. The AXIAL T 7lER I"0ALA::CE reduces the acuer level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced. The flux-to-flow ratio reduces the pcuer level trip ar.d associated reactor power-reactor power-imbalance boundaries by for a 15 flow reduction.

& J gtvand b.* } /.0S)

RC Pressure - Low, High and Pressure Tetoerature ,

The High and Low trips are provided to limit the pressure range in which reactor operatio.n is permitted.

During a sicw reactivity insertion startup accident frca low power or a slow reactivity inser: ion frca high power, the RC High Pressure setpoint is reached before the High Flux Trip Setpoint. The trip set-point for RC High Pressure, 2355 psig, has been established to maintain

- the system pressure below the safety limit, 2750 psig, for any design transient. The RC High Pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 2435 psia. The RC High Pressure trip also backs up the High Flux trip. -

The RC Low Pressure,1935 psig, and RC Pressure-Temperature (13.85 Tout *F-6498) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.32 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of D:B correlation limits, protecting against DNB.

Due to the calibration and instrumentation errors, the safety analysis used a RC Pressure-Temperature Trip Setpoint of ((13.85)

Tout *F-(6498)) psig.

High Flux / Number of Reactor Coolant Pumos On In conjunction with the Flux - a Flux-Flcw trip the High Flux / Number of Reactor Coolant Pumps On trip prevents the minimum core C:BR from.

decreasing belcw 1.32 by tripping the reactor due to the loss of reactor coolantpump(s). The pump monitors also restrict the power level for the number of pumps in operation.

~

JU?i :' i 25

,B 2-6 s DAVIS-BESSE, UNIT 1 I

i

.. _. - - . . .. . . . . . ._ _ _ .-. _ . . . _ _ . _ = . . .--- -- . -_ - . .

1 i t

1 i

1 l

> O, .

?

i i 4 t

l i

i

~

i

~

l l t . .

i .

I i

i s

THis P.t ' 5 DI M I~!'#' SE j ggpoj,,,, i M F D A WE M N #

ki o r reur l

i ,

i I 4

2 9 3

s .

t

, e e i

j BASES Figure 2.1 i

I

\ ' DAVIS-BESSE, UNIT 1 B 2-8 jut; ' .

' 1976 i

i I

i

1

. 5l l

p 3/4.1. REACTIVITY CONTROL SYSTE"5 ,

~

3/4.1.1 BOPATIO:1 CCNTROL l SHUTC:<:1 M*.2.3'N LIMITING CO'OITIO'l F00. 00ESATIO" 3.1.1.1 The SHUTD]WM MARGIN shall be 1 1% ak/k. -

. APPLICA3ILITY: M0]ES 1, 2*, 3, 4 and 5. ,

ACTION: .

With the SHUTDOWN MARGIN < 1% ak/k, immediately initiate and continue

. baration at > 13 gpa of 7575 ppr. boron or its equivalent, until the required SHJiCLN MARGIN is restored.

_ SURVEILLANCE REOUIREMENTS

( -

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 1 1% ak/k:

a. Within one hour after detection of an inoperable control r'od(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (.s) is inoperable. If the incperable control rod is ir=ovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the ire.ovable or untrippable control rod (s).

b. When in MODES 1 or 2 , at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, by verifying that regulating rod groups withdrawal is within the limits of Specification 3.1.3.6.

2

c. When in MODE 2"3 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criti-cality by verifying that the predicted critical control red position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL PC'.iER after each fuel leading by consideration of the fccters of e. below, with the regulating rod groups at the ::: wit.= 'm ert %rrlimit of Specification 3.1.3.6.  % gg

'With Kes.,, > 1.0.

' g's With Keff < l.0.

Sec Spccial Test Exception 3.10.4.

DAVIS-BESSE, UNIT 1 3/4 1-1 jut  : 1976

.I

REACTIVITY CONTRCL SYSTEMS

(~ MINIMUM TEMPERATURE FOR CRITICALITY LIMITIf;G CONDITIC'; FOR OPEP.ATION s

3.1.1.4 The Reactor Coolant System lowest loop temperature (Tavg) shall be 3,525'F.

APPLICABILITY: P.00ES 1 and 2*.

ACTION: a

< 525'F, restore '

• With a Reactor Coolant System loop temperature T H (Ttowithinitslini.twithi t$39next 15 minutes.

SURVEILLANCE RE0'JIREP.ENTS 4.1.1.4' The RCS temperature (Tayg) shall be determined to be 3,525'F: ~

a. liithin 15 minutes prior'to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is- critical and the Reactor Coolant System T avg islessthan,pNs.-F.

530*

s With K,ff 3,1.0.

%sa&= i & & s.m.a. .

N.

DAVIS-BESSE, UNIT 1 ,

3/4 1-5 Jgy p 3 797g a'

L

l REACTIVITY C0'iTDOL SYSTEMS 3/4.1.2 803."TICN SYSTEMS FLOW PATHS - SH'JTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 At least one of the following boron injection flow paths shall be OPERABLE.

a.

A flow path from the ::ne:ntrates boric acid ^ tor:;a systen "

via a boric acid pump and a rakeup or decay heat removal (DHR) ,

pump to the Reactor Coolant System.. if only the boric acid

- h_ tc --^ system in Specification 3.1.2.8a is OPERABLE, or

b. A flow path from the borated. water storage tank via a makeup or DHR pump to the Reactor Ccolant System if only the borated

- water storage tank.in Specification 3..l.2.Sb is OPERABLE.

APPLICABILITY: MODES S and 6. '

( ACTION: -

With none of the abcVe flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one i'njection path is restored to OPERABLE status.

SURVEILLANCE RECUIRE"ENTS -

4.1.2.1 At least one of the above required flow paths shall be demon-strated OPERABLE:

~

a. At least once per 7 days by verifying that the pipe tempera-b ture of the heat traced portion of the flow path is > 105'F when a flow path frca the concentrated boric acid storage system is used, and

b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position is in its correct position.

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

gg.g

~

. I, !

i REACTIVITY CONTROLS SYSTEMS 7 FLCW PATHS - OPERATING LIMITIt:3 CONDITIC': FOR OPEP.ATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERA 3LE:

cr W h'

a. A flow path frca the se -- 1 1 4 boric acid 34sc-;s system via a boric acid pump and cakeup or decay heat removal (CHR) pump to the Reactor Coolant System, and -

a

b. A flow path from the borated water storage tank via makeup or -

DHR pump to the Reactor Coolant System.

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

ACTION:

a. With the flow path from the concentrated boric acid s'torage system inoperable, restore the inoperable flow path to OPERABLE l

status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDSY and borated to a SHUTDONN MARGIN equivale'nt to 1% ok/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore tne flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within

.the nex,t 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the flow path frem the borated water storage tank in-operable, restore the flow path to OPERABLE status within one hour or be in at least HOT STAND 3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOCWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVIELLANCE REOUIRE"ENTS 1

4.1.2.2 Each of the above required flow paths shall be demonstrated OPERASLE:

a. At least once per 7 days by verifying that the pipe tempera-ture of the heat traced portion of the flow path from the

. concentrated boric acid storage system is t 105'F.

k DAVIS-BESSE, UNIT 1 3/4 1-7 JUN 3 51975 l

L .

1

REACTIVITY CO. TROL SYSTEMS g- DECAY HEAT REM 3 VAL PU'4P - SHUTD0'41 LIMITING C0!;DITIO!I FOR OPERATIO't 3.1.2.5 At least one decay heat removal (DHR) pump in the boron injection ficw path required by Specification 3.1.2.1 or 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE m r3er:rj bus.

APPLICABILITY: MODES 4,* 5* and 6.

^

ACTIO!1:

With no DHR pump OPERABLE, suspend all operations involving CORE ALTERATIOiS or positive reactivity changes until at least one DHR pump is restored

- to OPERABLE status.

~

SURVEILLAflCE REOUIREMENTS

(

4.1.2.5 No additional Surveillance Re'quirements other than those required by Specification 4.0.5. ,

' RCS Pressure < 300 psig.

e 6

. L ,

l

\. ,

DAVIS-BESSE, UNIT 1 -

3/4 1-11 JUN 2 5 574

a REACTIVITY C0!! TROL SYSTEMS

(~ ,

BORIC ACID PUMP - SHUTD0'a'*4 1

LIMIT 1f'G C0!!DITIOi FOR ODERATIO!!

~

3.1.2.6 At least one boric acid pump shall be OPERABLE and capable of being powered from an OPERABLE -[- ;^*cy bus if only the flow '

through the boric acid pump in Specification 3.1.2.la is CPERABLE.

APPLICABILITY: MODES 5 and 6. a ACTION:

With no boric acid pump OPERABLE as required to ccmplete the flow patti

.' of Specification 3.1.2.la, suspend all operations involving CORE ALTEPATIO!!S or positive reactivity changes until at least one boric acid pump is restored to 0?EPABLE status.

( .

SURVEILLAtCE RE0VIREMENTS 4.1.2.6 No additional Surveillance Requirements other than those required by Specification 4.0.5.

. i .

\.

(

DAVIS-SESSE, UNIT 1 .

3/4 1-12 gyg 2 _ .

0 .1375 L

- l,

i e

REACTIVITY CCN1ROL SYSTEMS -

(' BORIC ACID PUMPS - CPERATING e

LIMITING CONDITION FCR OPERATION 3.1.2.7 At least one boric acid pur in the boron injection flow path ~

required by Specification 3.1.2.2a all be OPERABLE and capable of being powered from an OPERABLE ere- sney bus if the flow path through the boric acid pump in Specification 3.1.2.2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With no boric acid pump OPERABLE, restore at least one boric acid pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTD0'a'N MAh^!N equivalent to 1% ak/k at 200*F within the next 6 ' sours; restore at least one boric acid pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.2.7 No additional Surveillance Requirements other than those required by Specification 4.0.5.

DAVIS-BESSE, UNIT 1 Jul,01 G76 3/4.1-13 l . s I

. t T. ,,,,, .-, t - m P.EACT!YITY CT:TP.0L SYSTEMS  ;, , ,. , . ; ,_. . j. _ . , .,

. , . r,,

1; P ,. .

6031TED L'ATER SOUR.~ES - ODERATING

' ' ' ' ' ~- . ' NI

__ _ u . ' - '" -

g. . . . . .

~~ --

LIMITiniC0"DITIO"F03.OPET.," TION 3.1.2.9 Each of the following borated water sourecs shall be OPERACLE: .

a. The concentrated boric acid storags system and as:oc'ated heat tracir.; with: '

l. A minicu:r. contained borated utter volume in accerc'ance 2 with Figure 3.1-1,

2. Betwecn 7875 and 12,125 ppm of boron, and

3. A mininum solution tacperature of 105'F.

. b. The barated water storage tank,(E"5T) with:

1. A contained borated water volume of between 402,500 and gallons,

( -

2. Between 1800 and ppm cf borod, and

3. ,

A minimum solution temperature of 35'F.

~

APPLICASILITY: MODES 1, 2, 3 and 4. .

ACTIO":

a. With the - -^mtr:.tsd boric acid at:r:10 system inopercble, restore the storage system to CPERACLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANCSY and borated to a SHUTD0'a'N mar $ GIN equivalent to 1S Ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore tha concentrated boric cccid storage system to OPERABLE status within the next 7 days or be in COLD SHUTD0'..'N within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the borated water storage tank inoperable, restore the tank to CPEP.ABLE status witnin or,e hcur or be in at least HOT

• STANDSY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'.!"

within the follcwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -

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

. JUL 0 g ;g,g

- 1 i_

^

Docket #

Control #bW Date Reevd. g' /7 - , .

DEFiti!TIO!;S Regulatory Docket File REPORTABLE OCCUR:.ENCE 1.7 A P.E:::.T:E'_E 0-~5:.E';;E shall be any of those conditions 3,gecified as a rep: t::le c:c rren:e in Revision 4.of Regulatory Guide 1.16,

" Reporting of Operating Infomation - Appendix "A" Technical Specifications."

CONTAir,ENT INTEGR'TY 1.C C0'iTAI4".EtiT IliTEGRITY shall exist when: . 2

a. All penetraticr.s required to be closed during acciderlt con- -

ditions are either: .

~

1. Capable of being closed by.- *llML' ~YOS __

r '

c_' _ . 9.=, or

b. Closed by r.anual valves, blind flanges, or deactivated autoT.atic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.1.

b. All equipment hatches are closed'and sealed,

c. Each airlock is OPER/.BLE pursuant to Specification 3.6.1.3, and

d. The contain :ent leakage rates are within the limits of Specification 3.6.1.2.

e. The sealing rechanisa associated with each penetration (e.g., ,.

welds, bellows or 0-rings) is OPERABLE.

CRANf4EL CALISPATIC'i 1.9 A CHAl;NEL CALISPATION shall be' the ( 'justrent, as 'necessary, of the channel output sucn that it respor.ds wnn necessary range and accuracy l to known values cf the paraceter which the channel in:nitors. The CHAN"El CALILRATICN shall encc: pass the entire channel including the senser and aisro and/or trip functions, r.d shall in:lude the CHANNEL FU:i:TIO:'AL TEST.

CHAti!.EL CALIEPAT:0N ray be performed by any series of sequential, over-lapping er total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessrent of channel behavior during operaticr. by observation. This detemination shall s

include,t.5ere possibl.2, ccrparison of the channel indication and/cr status with other indicatieris and/cr status derived frcm independent

, instrument channels r:easuring tne sare parcoeter.

DAVIS-BESSE, tri1T 1 1-2 JUL P 6 1" 4

/. .

DEFi:ITIC.S - -

ps ,

c. Reactor ccolant system leakage thrcugh a steam generator to the secor.dary sys tem.

UtilDE::TIFIED LEA!' AGE 1.15 U ;IDE::TIFIED LEAK *.GE shall be all leakage which is not IDENTIFIED LEATAGE or CC :TRCLLED LEAKAGE.

PRESSURE SO'";DARY LEAKAGE 1.16 PRESSURE EOU::DARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Ccolant System component body, pipe wall or vessel wall. ,

C0f;TP.0LLED LEAKAGE gg 1.i7 C0iTP0LLED LEAKAGE shall be that seal water flcu supplieo3 the reactor coolant pump seals. - -

QUADRA!!r PO* DER TILT -

Jgmd b the. Ecllown3 '9x? ion 1.18 OJADRA::T F0' DER TILT shall be the ;; ' y di'fe.c..w L::.... . the pew g:r.;rn;d i; ;g ;;r; , dc;nt ';;;;r ;r 1;..;r c;ce L .10 r.d W. a . u. ,; e ;;..a c c f al l ; ._ ic;r.t,; 'r th;t '.;1' D;;;. cr !m.;c' 3 he cera di.ided t., ;he c.5_;2 ;;..s cimil ; ..m.o .n in . .dli -

(-triner cr ic..er) cf th; c;c; and is exrressed in percent.

er in any core cuadrant '____. _. '.....-i

[ 100 s+*( Average po.;er of all quaorants M .  ;. . . .. . c ) , ) g '

DOSE EQUIVALEli 1-131 1.19 DOSE EQUIVALEt1T I-131 shall be that concentration of I-131 (pCi/ gram)

' hich a alone would produce the same thyroid dose as the quantity and isotopic nixture of I-131, I-132,1-133, I-134 and I-135 actusliy present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance factors for

)ower and Test Reactor Sites." -

- AVERAGE DISIITEGRATIO1 Ef;EP.GY l.20 E-AYERAGE DISI!1TEGRATIO:1 E:iERGY shall be the average (iteighted in 3roportion to the concentration of each radion'uclide in the reactor ccolant It the time of sampling) of the sum of the average beta and cc.wa energics iDAVIS-SESSE, UNIT 1 1-4 Jtly p 8 797c

[ .

l I

i DEFINITIONS l

f .

i per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 1S minutes, making up at least 95% of the total non-iodine activity in the coolant.

STAGGERED TEST BASIS  :

1.21 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains cr designated components obtained by dividing the specified test interval into n equal subintervals,

b. The testing of one system, subsystem, train or designated components at the beginning of each subinterval.

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is ~0PERABLE.

FREQUENCY NOTATION i 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in- Table 1.2.

AXIAL POUER IMBALANCE ,

1.23 AXIAL POUER IMBALANCE shal be the THERMAL POWER in the top half of the core expressed as a percentage of RATED THERMAL. POWER minus the THERMAL POWER in the bottom half of the core expressed as a percentage of RATED THEPJiAL POWER fe s. W. ..it.

SHIELD BUILDING INTEGRITY , ,

1.24 SHIELD BUILDING INTEGRITY shall, exist when:

a. The airtight doors and the blowout panels listed in Table 4.6-1 are closed except the airtight doors may be used for normal transit entry and exit. ,

b. The emergency ventilation system is OPERABLE. ,

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.2S The REACTOR PROTECTION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until power interruption at the control rod drive breakers-DAVIS-BESSE, UNIT 1 1-5

, . JUL 3 0197S 7,

4

- t "

h

t t, ..

i f .

! (

4 i +

J j .

l ,

?

i . .. .

' i i

e 1'

1 4

THis F- .:'

Ec;  ;< P :.3 RECE PT OF l f

! *l; lNf05o. IU: I OM T r.E A;-PLICAMI u nf Tro F i .

i

i '

i-4 i . .

I i l .

I  !

! t i .- Figure 2.1-1

! Reactor Core Safety Limit -

?

' DAVIS-BESSE, UNIT 1 2-2 JUN 2 41976

).