Text

Amend 33 to License NPF-3 Modifying Tech Specs to Permit Operation for Second Cycle & to Reflect Changes in RCS High Pressure & Power Operated Relief Valve Setpoints
	ML19340B075
Person / Time
Site:	Davis Besse
Issue date:	10/01/1980
From:	Reid R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19340B076	List: ML19340B075; ML19340B077; ML19340B078;
References
	NUDOCS 8010210090
	Download: ML19340B075 (58)
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NUCLEAR REGULATORY ccMMISSION

=4 manaron.a.c.mossa f

THE TOLEDO EDISON COMPANY

..a AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 1

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 AMEN 0 MENT TO FACILITY OPERATING LICENSE handment No. 33 l

License No. NPF.3 i

1.

The Nuclear Regulatory Conmission (the Consission) has found that:

4 A.

The applications for amendment by the Toledo Edison Company ane i

The Cleveland Electric Illuminating Company (the licensees) 4 dated February 11, 1980, as revised and supplemented, and July 13, 1979, comply with the standards and requirements of the Atomic Energy Act of 1954, as arnended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 3.

The faci 11ty will operate in confomity with the applications, the provisions of the Act, and the rules and regulations of the Connission; i

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health j

and safety of the public, and (ii) that such activities will be

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conducted in compliance with the Consission's regulations; D.

The issuance of this amendnent will not be inimical to the connon defense and security or to the health and safety of the public; and j

E.

The issuance of this amendment is in accordance with 10 CPR Part 51 of the Connission's regulations and all applicable requirements have been satisfied.

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i 2.

Accordingly, Facili.ty Operating License No. NPF-3 is hereby amended as indicated below and by changes to the Technical Specifications as, indicated in the attachment to this license amendment:

A.

Revise paragraph 2.C.(2) to read as follows:

Technical Specifications l

The Technical Specifications contained in Appendices A and B, j

as revised through Amendment No. 33, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

B.

Delete paragraphs 2.C.(3)(e), 2.C(3)(f) and 2.C.(3)(g).

3.

This license amendment is effective as of the date of its issuance.

i FOR THE NUCLEAR REGULATORY COMMISSION h. ' ]L l.

l Rouert W. Reid, Chief Operating Reactors Branch !4 Division of Licensing

Attachment:

Changes to the Technical l

Specifications Date of Issuance: October 1,1930 b

g

I i

l ATTACHMENT TO LICENSE AMENDMENT NO. 33 FACILITY OPERATING LICENSE NO. NPF-3 i

DOCKET NO. 50-346 1

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Pages 3/41-35 (New Page) 2-2 3/4 1-36 2-3 3/4 1-37 2-5 3;: 3 33 2-7 3/4 2-1 2-8 3/4 2-2 B 2-1 3/4 2-2a B 2-2 3/4 2-3 B 2-3 it 3/4 2-3a i

B 2-5 3/4 2-4 B 2-6 3/4 2-4a B 2-8 3/4 2-12 3/4 1-26 3/4 2-14 3/4 1-28 3/4 4 1 3/4 1-28a 1

3/4 4-4 1

3/4 1-29 3/4 10-1 3/4 1-29a 3/4 10 2 3/4 1-29b 8 3/4 1-2 3/4 1-29c B 3/4 1-4 3/4 1-31 8 3/4 2-1 3/4 1-34 (New Page)

B 3/4 2-3 B 3/4 4-1 B 3/4 4-la

hk 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The ecmbination of tne reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANCBY within one hour.

REACTOR CORE 2.1.2 The combination of reactor THEMAL POWER and AXIAL POWER IMSALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of two, three and four reactor coolant pump operation.

Appl!CASILITY: M00E 1.

ACTION:

Whenever the point defined by the combination of Reactor Coolanc System flow, AXIAL POWER IMBALANCE and THE.:tMAL POWER has exceeded tt.e appropriate safety limit, be in HOT STANC8Y within one hour, REACTOR COOLANT SYS~EM PRESSURE 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig.

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

ACTION:

MODES 1 and 2 - Whenever the Reactor Coolant System pressure has ex-caeded 2750 psig, be in HOT STANC8Y with the Reactor Coolant System pressure within its limit within one hour.

M00E3 3, a

- Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

DAVIS-BESSE, UNIT 1 2-1 I

2tiOO L.

i RC High Pressure Trip 619.0,2300 RC Higa 2200 Temperature Trip 6 19. O, 21130 ACCIPTA3LE 2

OPERATION Safety Limit S

RC Pressure Temperature Trip 2000 606.7,1985 a.

RC Low Pressure Trip l

f i

1800 l

l 590 600 GIO 620 630 f,

Reactor Outlet Temperature,*F FIGURE 2.1-1 Reacco: Cars Safety LL:1:

i 2AV!S-3ISSE, '.RCT I 2-- 2 Amendment Plo. I6 3 3

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FOR SPECIFIED RC PUNP

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COM8IMITION

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UNACCI?!A3L2 QtACOZPTABLE

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OPERATION OPERATION I

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-20 0

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i FI GURE 2.1-2 Rese:or Core. Safety L* 4:

PUMPS OPERATlHG REACTOR CCOLANT FLOW (GPM) 4 387.200 3

190,i00 i

Or.IS-3 Esse, CiIT 1 2-3 Amendment No.J1,16, 3 3 l

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS 1

REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

1 APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Protection System instrumentation setpoint less conserv-ative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

m DAVIS-BESSE, UNIT 1 24

TABLE 2.2-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPolNT ALLOWADLE VALUES

[

1.

Manual Reactor Trip Not Applicable Not Applicable E

2.

liigh Flux

< 105.5% of RATED TilERMAL POWER

< 105.6% of RATED TilERMAL POWER M

Eith four pumps operating Ulth four pumps operating #

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< 80.3%of RATED THERMAL POWER

< 80.2% of RATED TilERMAL POWER With three pumps operating Eith three pumps operating #

3.

RC liigh Temperature

< 619"F

< 619.08"F I

7 4.

Flux - A Flux-Flow Trip Setpoint not to Allowable Values not to exceed exceed the limit line of the limit line of Figure 2.2-2.#

figure 2.2-1.

UI 5.

RC Low Pressure 1 1985 psig 1 1984.0 psig*

1 1976.5 psig**

6.

RC liigh Pressure 1 2300 psig 1 2301.0 psig*

1 2308.5 psig**

UI I

7.

RC Pressure-Temperature L (12.60 T F - 5660) psig g (12.60 T F - 5660.41) psig out at l'

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\\35.0,76.0 setpoint for a 25f, flow reduction THREE PUMP for three pump LlHIT LINES operation. The

~~50 actual trip sr.cpoint:

-M. 0, 57. 2 will be cal.ulated by the RP". and will

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tional to the actual

- g flow with three pumps.

.. 20 ACCEPTABLE OPERATION FOR UNACCEPTABLE UNACCEPTABLE SPECIFIED RC PUNP COMBINAT!0N OPERATION QPERATION

-60

-40

- 20 0

20 40 GO Axial Power imbalance, 7 FIGURE 2.2-I

1p seepoinc for Flux-1 Flux-Flow A" 3-3 Esse, UNIT 1 2-7 Amendment No. M. JI,33

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FIGURE 2.2-2 Allevable value for Flux,a Flux-Flov DA7'I-3ISSE, UNIT 1 2-3 Amendment No.,1(,1( 3 3

I 2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel I

cladding and possible cladding perforation which would result in the

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relea.se of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction 17 heat transfer coefficient. CNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the B&W-2 DNB corr-lation. The DNB correlation has been developed *w predict the DNB flux and the location of DNB for axially unifom and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that 'nould cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational trans'ents, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

1 The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 387, 200 GPM, which is 110% of i

design flow rate for four operating reac*ar coolant pumps. This curve is based on the following hot channel factors with potential fuel densif1-I cation and fuel rod bowing effects:

)

FQ = 2.56; F.H = 1.71; F)=1.50 N

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 withdrawal, and fom the core DNBR design basis.

DAVIS-BESSE, UNIT 1 32-1 Amendment No.11I3 3

SAFETY LIMITS BASES The reactor trip envelope appears to approach 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 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thennal limits and account for the effects of potential fuel densification and potential fuel rod bow:

1.

The 1.30 DNBR limit produced by a nuclear power peaking l

factor of Fq = 2.56 or the combination of the radial peak, axial peak and position of the axial peak that yields no less l

than a 1.30 DNBR.

I 2.

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

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

The specified flow.-ates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thennal power combinations shown in SASES Figure 2.1.

The curve of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible l

thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to +22"., whichever condition is more restrictive. This curve includes the potential effects of fuel rod bow and fuel densification.

The DNBR as calculated by the B&W-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher. Extrapolation of the correlation beyond its published quality range of +22". is justified on the basis of experimental data.

DAVIS-BESSE, UNIT 1 B 2-2 Amendment No.,W, 3 3

SAFETY LIMITS BASES For the curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than +22%

for that particular reactor coolant pump situation. The 1.30 DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four pump curve will be above and to the left of the three pump curve.

2.1. 3 REACTOR CCOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the mtegrity of the Reactor Coolant System from overpressurization and therwy prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section

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III of the ASME Boiler and Pressure Vessel Code wh'Oh permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor Caolant System piping, valves and fittings, are designed to ANSI B 31.7, 1968 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component 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 OAVIS-BESSE, UNIT I B 2-3 Anendment No. 4 3 3

.2, 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 4

2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETp0INTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints ha.e bien selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip setpcint less conservative than its Trip Setpoint but within its scecified Allowable Value is accept-acle on the basis that each Allowable Value is ecual to or less than the drift allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certain functier.s of the Reactor Protection System in order to permit control rod drive tests, zero power FHYSICS TESTS and certain startuo and shutdown procedures.

The purpose of the Shutdown Bypass High Pressure trip is to prevent j

normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The Hign Flux Trip Setpoint of < 5.0% prevents any significant reactor power i

from being procuced. fufficient natural circulation would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumos were operating.

Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic 3

Reactor Protection System instrumentation channels and provides manual reactor trip capability.

Hich Flux A High Flux trip at high power level (neutron flux) provides reactor core proteccion against reactivity excursions which are too rapid to De protected by temoerature and pressure crotective circuitry.

Ouring normal station operation, reactor trip is initiated when the reactor power level reaches 105.5% of rated power. Due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, wnich was used in the safety analysis.

DAVIS-BESSE, UNIT 1 324 l

LIMITING SAFETY SYSTEM SETTINGS BASES RC Hiah Temaerature The RC High Temperature trip $_ 619'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.

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

The power level trip setpoint produed by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.

The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum pennissible power level, and for every power level there is a minimum permissible low flow rate. Examples of typical power level and low flow rate continations for the pump situations of Table 2.2-1 that would result in a trip are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if power is 107% and reactor coolant flow rate is 100% of full flow rate, or flow rate is 93.3% of full flow rate and power level is 100%.

2.

Trip would occur when three reactor coolant pumps are operating if power is 80.2% and reactor coolant flow rate is 74.9% of full flow rate, or flow rate is 69.8% of full flow rate and power is 75%.

For safety calculations the maximum calibration and instrumentation errors for the power level were used. Full flow rate in the above two examples is defined as the flow calculated by the heat balance at 100%

power.

DAVIS-BESSE, UNIT 1 B 2-5

/v endment No. !.16', 3 3

LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE 'oundaries are established in order to c

prevent reactor thermal. limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the powe level trip produced by a flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

RC Pressure - Low, Hioh and Pressure Temoerature The High and Low trips are provided to limit the pressure range in i

which reactor operation is permitted.

During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RC High Pressure j

setpoint is reached before the High Flux Trip Setpoint. The trip set-point for RC High Pressure, 2300 psig, has been established to maintain l

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 psig. The RC High Pressure trip also backs up the High Flux trip.

The RC Low Pressure,1985 psig, and RC Pressure-Temperature (12.60 T

F-5650) psig, Trip Setpoints have been established to maintain the 08Etratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction.

It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against ONB.

High Flux / Number of Reactor Coolant Pumos On In conjunction with the Flux - 1 Flux-Flow trip the High Flux / Number of Reactor Coolant Pumps On trip prevents the minimum core ONBR from decreasing below 1.30 by tripping the reactor due to the loss of reactor l

coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

DAVIS-BESSE, UNIT I B 2-6 Amendment No. 3 3

.,4 4

LIMITING SAFETY SYSTEM SETTINGS BASES Containment Hich Pressure The Containment High Pressure Trip Setpoint < 4 psig, provides positive assuranca that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RC Low Pressure trip.

t 4

BAVIS-BESSE, UNIT I B 2-7

A d

2400 2300

^

3 S

2200 5

E 3

2100 e

2 C"

2000 1900 1800 s

1700 580 590 600 61 0 SID S30 Resctor Outlet Temperature -(*F)

RC FLOW POWER 387,200 GPW 1125 Pressure /Te=perature L1=1ts at Max 1=u=

Allowable ?Over fc Mini =u= DN3R BASES FIGURE 2.1 lu s-SEssI, U m 1 3 2-8 Amendment No E,3 3

s REACTIVITY CONTROL SYSTEMS SAFETY 't00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All safety rods shall be fully withdrawn.

APPLICABILITY:

l* and 2*f.

ACTION:

With a maximum of one safety red not fully withdrawn, except for sur-veillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a.

Fully withdraw the red or b.

Declare the red to be inoperable and apply Specification 3.1.3.1.

SURVE!LLANCE REOUIREVENTS 4.1.3.5 Each safety rod shall be detemined to be fully withdrawn:

a.

Within 15 minutes prior to withdrawal of any regulating rod during an approach to reactor criticality.

b.

At least once per 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months thereafter.

'See speclai Test Exceptica 3.10.1 and 3.10.2.

(With K,ff > 1.0.

DAVIS-BESSE, UNIT 1 3/4 1 25

7 i

REACTIVITY CONTROL SYSTEMS REGULATING RCD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.l.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a and -2b and 3.1-3a and -3b, with a_ rod group overlap of_ 25 + 5% between secuential withdrawn groups 5, 6, and 7.

APPLICABILITY: MODES 1* and 2*f.

ACTION:

With the regulating rod groups inserted beyond the above insertion limits (in a region other than acceptable operation), or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

I a.

Restore the regulating groups to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or c.

Be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

NOTE:

If in unacceptable region, also see Section 3/4.

'See special Test Exceptions 3.10.1 and 3.10.2.

With Keff > 1.0.

DAVIS-BESSE, UNIT 1 3/4 1-26 Amendment No.e W, 3 3

l REACTIVITY CONTROL SYSTEMS REGULATING R00 INSERTION LIMITS SURVEILLANCE REQUIREE NTS 4.1.3.5 The position of each regulating group shall be determined to be within the insertion, sequence and overlap limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when:

a.

The regulating rod insertion limit alarm is inoperable, then verify the groups to be within the insertion limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; b.

The control rod orive sequence alarm is inoperable, then verify the groups to be within the sequence and overlap limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

DAVIS-BESSE, UNIT 1 3/41 27

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i 5$

5 l

y>

g (2uo.5.10 (500,102) en 100 1

(260)ool Power Level

+Cutof f 92%

74.821 l

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4 g

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TEST TARGET (MT-3) 4 1.0 lnm su i

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Y CAVIS-BESSE, UNIT 1 3/4 1-29b Amendment No. M 3 3

-_________o______

Figure 3.1-3d Deleted j

l l

l i

DAVIS-BESSE, UNIT 1 3/4 1-29c Amendment No.,1{ 3 3

l REACTIVITY CONTROL SYST245 R00 PROGRAM LIMITING CONDITION FOR OPERATION 3.1.3.7 Each control *od (safety, regulating and APSR) shall be pro-grannec to acerate in the core position and rod group specified in Figure 3.1-4 l

APPLICABILITY: MODES l' and 2*.

ACTION:

With any control rod not programed to coerate as specified above, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

J SURVEILLANCE REQUIREMENTS 4.1.3.7 a.

Each centrol red shall be demonstrated to be programed to operate in the specified c:re position and rod group by:

1.

Selection arm actuation fr::m the centrol room and verifi-cation of rt.9ement of the precer red as indicated by both the absolute and relativa position indicators:

a)

For all control reds, after the control red drive patenes are locked subsequent to test, rcprograming or maintenance within the panels.

b)

For specifically affected individual rods, follcwing

{

maintenance, test, reconnection or modification of power or instrumentation caules from the c:ntrol rod drive centrol system to the control red drive.

2.

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

b.

At least once each 7 days, verify that the control rod drive I

paten panels are locked.

'See special Tes: Exceptions 3.10.1 and 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-30 Amendment No.ll i

+

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3 to 11 12 13 14 15 No. af Group Roos

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4 Safety 3

5

!afety 4

1

!afety 3

Oontrol S

8 Control 7

12 Oontral i

1 APS1ts Total St CGURI 3.1-4 Con::a1 Red Cars *.aca-

1ons and Group Assigs-nents - Davis-Bessa 1 Cycla 2 DAVIS-BESSE, UNIT 1 3/4 1-31 Amendment No.,W, 3 3

A a

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4 1

2 I

INTENTIONALLY LEFT BLANK 1

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1 1

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1 OAVIS-BESSE, UNIT 1 3/4 1-32 Amendment No.11 i

I i

f

REACTIVITY CONTROL SYSTEMS XENON REACTIVITY LIMITING CONDITION FOR OPERATION 3.1.3.8 THERMAL POWER shall not be increased above the power level cutoff specified in Figure 3.1-2 unless one of the following conditions is sa tisfied:

a.

Xenon reactivity is within 10 percent of the equilibrium value for RATED THERMAL POWER and is approaching stability, or b.

THERMAL POWER has been within a range of 87 to 92 percent of RATED THERMAL POWER for a period exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in the soluble poison control mode, excluding xenon free start-ups.

\\

APPLICABILITY: MODE 1.

ACTION:

With the requirements of the above specification not satisfied, reduce THERMAL POWER to less than or equal to the power level cutoff within 15 minutes.

SURVEILLANCE REOUIREMENTS 4.1.3.8 Xenon reactivity shall be determined to be within 10% of the equilibrium value for RATED THERMAL POWER and to be approaching stability or it snall be determined that the THERMAL POWER has been in the range of 87 to 92% of RATED THERMAL POWER for > 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , prior to increasing THERMAL POWER above the power level cutoff.

DAVIS-BESSE, UNIT 1 3/4 1-33

REACTIVITY CONTROL SYSTES AXIAL POWER SHAPUG RCD INSER* ION LDiITS L31 TUG CONDICON FOR OPERATION 3.1.3.9 The axial power shaping red group shall b 11mized in physical inser-tion as shown on Tigures 3.1-Sa, -5b, -5c, and -5d.

i APPLICA3ILITY: If0 DES 1 and 2*.

ACTION:

With the axial power shaping rod group outside the above insertion 11 sits, I

either:

Restore the axial power shaping rod group to within the limits a.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or i

b.

Reduce THE'tfAL POWER to less than or equal to that fraction of RATED THE'CfAL POWER which is allowed by the red group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or i

c.

Be in at least HOT STXiDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1 1

SURVIILLANCE REOUIREfENTS 4.1.3.9 The position of the swal power shaping rod group shall be deter-mined to be within the insertion 14-4ts ac least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

. 1 1.0.

Wi:h K erf DAVIS-BESSE, UNIT 1 3/4 1-34 Amencment No. 3 3

pgj3oElk (12.102)

(30.102) 100 (8,92)

(35,92)

O 3

80 (f(3,80)

(40,80) a.

3 RESTRICTED REGION O5 E

SD i

c=

d (5,50)

(54,50) f a

(100,45

=

S 40 5

t a

PERMISSIBLE I

a.

OPERATING 20 l

REGION f(0,0) 0 0

20 40 50 80 100 l

APSR Position (", witnarawn) i i

I FIGURE 3.1 5a APSR Posicion L1=1:s, O to 150 i o gy?9,

(

l Tour RCPs -

Davis-Besse 1, Cycle 2

,i i

DN/IS-3E552, 'OC 1 3/' 1-35 Amendment No. 3 3

4 fIh x

(12,102)

(30,t02) e 100 0

(8,92)

(35,92)

E E

80 (5,80)

(40.80)

=

5_

RESTRICTED

~

REGION 3

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=

g o E

(54.50) 9 (,..s0) l

=

3 t

(100.45) 7 D

40 O

E PERMISSIBLE s.

OPERATING REGION 20 i

I h

I 0

0 20 40 60 80 100 APSR Position (5 #itnarawn)

FIGURE 3.1 5 h A?SR Posi:1on Limits, Af:er 150 *10 EF?D, Four RC?s - Davis-3 esse 1, Cycle 2 DAVIS-3 ESSE,

'v" TIT 1 3/4 1-36 Amendment No. 3 3

i 100 A

j 80

- (12. 77)

(30,77)

E 5

(8,69.5)

(35,59.5)

E RESTRICTED.

[

50 (8,50.5)

(40,60.5)

REG!0N E

E 40 (54,38)

S (5.38)

(100,34.25) 7 0

a 20 PERMISSi gl.E OPERATING REGION

( 0,0 )

a i

0 20 40 60 80 100 APSR Position (5 Witncrawn) l FIGURE 3.1 -5 e APSR ?ositics I.imi:s, O to 150 t o gypp, i

Three RC?s - Davis-3 esse 1, Cpcle 2 DAVIS-3 ESSE, t!N!! 1 3/a 1-37 Amendment

.33

D**3D

]D

' T Y /A ooN w in.1. Nli%

100 n

3 80

- (12,77)

(30.77) 2 3

(8.59.5)

(35.59.5)

=

SD (5.50.5)

(40,50.5)

RESTRICTED

=

3 REGION E

l

'E l

E 40 (64.33)

S (5,38)

(100,24.25)

S i

~

l O

=f PERMI SS 151.E 20 OPERATING REGION (0,0) l 0

0 20 40 SO 80 100 l

APSF Position (5 #itnarawn)

FIGURE 3.1 -5 d APSR Posd.:ics 7 <-1::s, After 150 t o I??D, i

Thraa RC?s - Davis-3 esse 1, Cycle 2 ii l

l l

JN/IS-3 ESSE, CiIT I 3/4 1-38 Amendment No. 3 3 l

l

k 3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1 and 3.2-2.

APPLICABILIN: MODE I above 40t of RATED THERMAL POWER.*

ACTION:

With AX!AL POWER IMBALANCE exceeding the limits specified above, either:

a.

Restore the AXIAL POWER IMBALANCE to within its limits within IE minutes, or b.

Se in at least HOT STANDBY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REOUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above act of RATED THERMAL POWER exceot when the AXIAL PCWER IMBALANCE alarm is inocersble, then calculate the AXIAL POWER IMSALANCE at 1. east once per hour.

See Spectai Test Exceptien 3.10.1

[

OAVIS-BESSE, UNIT 1 3/4 2 1 Amendment No. 3 3 l

l

g**S

.b k n

D D'

If

)

ao a

(-14.5,102)

(+10,102)

(-16,92)

(+12,92) n

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(+20,80) 5 m.

~

5 E

~

60 --

?,

~

e ( 30,50)

(*25.50)

O 40 -

5 i

==

PERMISSIBLE t

OPERATIN3 l

l 1

REGION 2

20 - -

i I

i i

30

-20

-10 0

+10

' 20

+35

+

Axial Pansr imoalanca (5) i FIGURE 3.2 -t a Axial ?:ver I=halance ?isi:s, O to 150 t10 IF?D, Four IC?s - Davis-3 essa 1, Cycle 2 OA*CS-3 ESSE, UNC 1 3/l. 2-2 Amendment No. Z 3 3

[

l u

-*~-*---e==.-.

we<

m.

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(-18.2,102) r m (+15,102)

(-20,92)

(+15,92)

RESTRICTED REGION 1

i

( 27,80) 80 3

(+20,90) a.

e#

60 ~

3

=

t (+25,50)

(-35.50) 3 40-5 u

2_.

1 4

3 B

PERMISS181.E 5

20-OPERATING t

REGION

.i i

30

-20

-10 0

+10

+20

+30 Axial Power I:::catance (5)

FIGURE 3.2 1b Axial Power Imbalance ? *~d:s, After 150 t10 22?D, Four RC?s - Davis-3 esse 1, Cycle 2 i

Dri!S-3ESSI, CTIT 1 3/4 2-2a Amendment No. X 3 3

a s

o::A:RONA2 100 --

80 -

RESTRICTED

(-10972)

(+7.5,77)

REGION

( -12,59. 5)

-(+9. 5 9. 5) 5 3.

1

( -15. 1. E0. 5 )

+15, 50. 5) 50 - -

5 2

5

( 22.5.28) 40 - *

(+18..

33) n,

(

i 5

S E

PERMISSIBLE 20 -

{

OPERATING g

s REGION l',

i

-30

-20

-10 0

+10

+20

+30 Azial Power im::alance (",)

FIGURE 3.2-2a 1

Axial Power Imbalance L*-4:s, O :o 150 t10 E3?D, Three RC?s - Davis-Besse 1, Cycle 2 OA7!S-3 ESSE, UN'd 1 3/4 2-3 Amendment No. E, 3 3 9

J

f. of Rated Thermal Power

. 100 RESTRICTED REGION

(-13.7,77)

. 80 (g3,3,77)

(-15,69.5)

(lI.3,69.5)

I

(-20.3,60.5)

-- 60 (l5,60.5) e

(-25. 3. 38)

(18.75,38)

.. u.o P ERMI SSI!!.E OP ERATING REGION

~ 20 t

9 f

e

-30

-20

-10 0

10 20 30 Axial Power imbalance, ",

FIGURE 3.2-2b Axial Power L: balance Li:si::s, Afrar 130 Il0 EFFD,

" Area RC?s - Davis-Besse 1, Cy:'.e 2 l

DA7 5-3 ESSE, *.iNIT 1 3/4 2-3a Amendment No. X, 3 3 l

4 i

i Figure 3.2-3b Deleted i

i t

f DAVIS-BESSE, UNIT 1 3/4 2 Ja Amendment No. }i', 3 3

i 1

POWER DISTRIBUTION LIMITS LIMITING CONDITICN FOR OPERATION (Continued)

ACTION:

(Continued) d.

With the QUA05' ANT POWER TILT determined to exceed the Maximum Limit of Table 3.2-2, rMuce THERMAL POWER to < 15" of RATED THERMAL POWER withi". noces.

SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT shall be determined to be within the limits at least once every 7 days during oceration above 15". of RATED DiERMAL POWER except when the CUADRANT POWER TILT alam is incoerable, then the QUADRANT POWER TILT snall be calculated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l d*

CAVIS-BESSE, UNIT 1 3/4 2-11

e n

s N

TABLE 3.2-2 QUADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT Measurement Independent QUADRANT POWER TILT 4.92 11.07 20.0 QUADRANT POWER TILT as j

Measured by:

Symmetrical Incore Detector System 3.21 8.71 20.0 l

Pcwer Range Channels 1.96 6.96 20.0 Minimum Incore Detecto-System 1.90 4.40 20.0 l

1 DAVIS-BESSE, UNIT 1 3/4 2-12 Amendment No. X, 3 3

2 s

o POWER DISTRIBUTION LIMITS ONB PARAMETERS I

LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Reactor Coolant Hot Leg Temperature.

b.

Reactor Coolant Pressure c.

Reactor Coolant Flow Rate APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the param-eter to within its. limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE0VIREME'4TS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

DAVIS-BESSE, UNIT 1 3/4 2-13

TABLE 3.2-1 o!!

DNB MARGIN i2 ja LIMITS U

N?

Four Reactt,.-

Three Reactor Coolant Pumps Coolant Pumps E

Parameter Operating Operating il Reactor Coolant ilot Leg

< 610

< 610III

_o Temperature T *F g

Reactor Coolant Pressure, psig.(2)

> 2062.7

> 2058.7 i

III II Reactor Coolant flow Rate, gpm

> 396,880

> 297,340 1

R u

E Applicai,le to the loop with 2 Reactor Coo'lant Pumps Operating.

(2) Limit not applicable during either a filERMAL POWER ramp increase in excess of 5% of af RATE 0 TilERMAL POWER per minute or a illERMAL POWER step increase of greater than 10%

of RATED TilERMAL POWER.

g (3)These flows include a flow rate uncertainty of 2.5%.

N 4.

!s C4 r

9 9

3/4.4 REACTOR COOLANT SYSTEM 4

REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION

)

3. 5.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.*

ACTION:

)

MODES 1 and 2:

a.

With one reactor coolant pump not in operation, STARTUP and POWER OPERATION may be initiated and may ;:roceed provided THERMAL POWER is restricted to less than 80.2". of RATED l

THERMAL POWER and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the setpoints for the following trips have been reduced to the values specified l

'n Specification 2.2.1 for operation with three reactor coolant pumps operating:

1.

High Flux 2.

Flux-aFlux-Flow i

j "See Special Test Exception 3.10.3.

DAVIS-BESSE, UNIT 1 3/4 4-1 Amendment No. X 33

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o 9

REACTOR C001 ANT p,$1EM LIMITING C'ONDITION FOR CPERATION (Continued)

MODES 3, 4 and 5:

Operation may rareceed previded at least one reactor coolant loop a.

is in operation with an associated reactor coolant pump or decay heat removal pump.'

b.

Not more t*an one decay heat removal pumo may be operated with the sole suction path througn OH-11 and DH-12 unless the control power has been removed from the OH-ll and OH-12 valve operators, 4

or manual valves OH-21 and DH-23 are ocened.

C-The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

' Ail reactor coolant pumos and decay heat removal train pumos may be de-energi:ec 'or uc to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to I::orrodate surveillance testing

& ::re-ocerational testing, provided no coerations are permitted wnich could cause dilution of the reactor coolant system baron Concentratien.

i i

1 SURVE!!. LANCE RECU! RESENTS d

i 4.4.1 The ' Reactor protective Instrumentatien channels specified in the 1

apolicaole ACTION statement above shall be verified to have had their trip setpoints changed to tne values soecified in Soecification 2.2.1 for the

)

acclicable num::er of reactor coolant pumos operating eitner:

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a different pump combination a.

if the switch is made while operating, or

{

j b.

Prior to reactor criticality if the switch is made while shutdown.

DAVIS-BESSE, UNIT 1 3/44-2 Anendment No. 7, I 2 8 -

'UL 25 *.3e3

r

~

. i-REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2435 PSIG + 1%.*

i i

APPLICABILITY: MODES 4 and E.

ACTION:

With no pressurizer code safety valve OPERABLE, imediately suspend all i

operations involving positive reactivity changes and place an OPERABLE OHR loop into operation in the shutdown cooling mode.

't SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required j

by Specification 4.0.S.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

)

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

4 REACTOR COOLANT SYSTEM SAFETY VALVES AND ELECTROMATIC RELIEF VALVE - OPERATING i

LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2435 PSIG i 1,

When not isolated, the pressurizer.

electromatic relief valve shall have a trip setpoint of > 2390 PSIG and an allowable va'ue of a 2385.5 PSIG.**

APPLICABILITY: MODES 1, 2 and 3.

i ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutas or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 i

4 sSURVEILLANCE RECUIREMENTS 4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5.

For the pressurizer electrematic relief valve a channel cali-i bration check shall be performed every 18 months.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

.** Allowable value for channel calibration check.

,,0 AVIS-BESSE, UNIT 1 3/4 4-4 Amendment No. 3 3

o 3/4.10 SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.1 The group height, insertion and power distribution limits of i

g Specifications 3.1.3.1, 3.1.3.2, 3.1. 3. 5, 3.1. 3. 6, 3.1. 3. 7, 3.1.3.9, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

)

a.

The THERMAL POW 2R is maintained 3,85". of RATED THERMAL POWER, 4

b.

The High Flux Trip Setpoint is 5,101 of RATED THERMAL POWER higher than the THERMAL POWER at which the test is performed, with a maximum setting of 90% of RATED THERMAL POWER, and c.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in 4.10.1.2 below.

APPLICABILITY: MODE 1.

ACTI0f :

f With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1. 3.1, 3.1. 3.2, 3.1. 3. 5, 3.1.3.6, 3.1.3.7, 3.1.3.9, 3.2.1 or 3.2.4 are suspended, either:

a.

. Reduce THERMAL POWER sufficiently to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or b.

Be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS I

4.10.1.1 The High Flux Trip Setpoint shall be determined to be set within the limits specified within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initiation of and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during PHYSICS TESTS.

4.10.1.2 The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at least once per two hours during PHYSICS TESTS.

4

'I DAVIS-BESSE, UNIT 1

-3/4 10-1 Amendment No. 3 3

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.2 The limitations of Specifications 3.1.1.3, 3.1. 3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, and 3.1.3.9 may be suspended during' the perfo'rmance of PHYSICS TESTS provided:

a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints on the OPERABLE High Flux Channels are set at i 25% of RATED THERMAL POWER.

c.

The nuclear instrumentation Source Range and Intermediate Range high startup rate control rod withdrawal inhibit are OPERABLE.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 5% of RATED THERMAL P0kER, immediately open the control rod drive trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 Each Source and Intermediate Range and High Flux Channel shall be subjected to a CHANNEL FUNCTIONAL TEST withir.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

DAVIS-BESSE. UNIT I 3/4 10-2 Amendment No. 33

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4 m'

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oo o

o 3/4.1 REACTIVITY CONTROL SYSTEMS SASES 1/a.1.1 BORATION CONTROL 1

3/4.1.1.1 SHUTDOWN MARGIN l

A sufficient SHUTDCWN MARGIN ensures that 1) the reactor can be ude subcritical free all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reac:ce will be maintained sufficiently subcritical to preclude inadvertent criticality in :Me shutdown condition.

During Modes 1 and 2 the SHUTCCWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion l

limits.

J SHUTCCWN MARGIN requirements vary throughout core lift as a function i

of 'uel decletion, RCS boren concentration and RCS T The most restrictive candition occurs at ECL, with I at no Ye,ad coerating a

temcerature. De SHUTCCWN MARGIN required N9consisten: with FSAR safety analysis assumotions.

i 3/a.T.1.2 BORON DILUTION A minimum ficw rate of at least 2S00 GFM crovides adecuate mixing, j

prevents stratification and ensures that reactivity enanges will be gradual thrcugn the Reactor Ccolant System in :ne core during bcren concentration reductions in the Reactor Coolant System. A '1cw rate of 4t least 2SCO GPM will circulate an ecuivalent Reacter Ccolant System volume of 12,110 cuoic feet in accroxima:ely 30 minutes.

The reactivity

]

cnange rate associated with boren concentration reduction will be within

)

the cacacility for coerator recognition and control.

3/4.1.1. 3 MODERATOR TEMPERA ~1JRE COEF ICIENT i

.he limitations on moderator temperature coefficient (MTC) are i

provided to ensure that the assumotions used in :ne accident and transient analyses remain valid through each fuel cycle.

The surseillance recuire-ment for measurement of ne MTC each fuel cycle are acecuate to confirm

ne MTC value since this coefficient changes sicwly due ;rinci ally c One reduction in RCS baron concentration associated with fuel burnuc.

i ne confir-nation that the measured EC value is within its limit provides assurance that the coefficient will ce maintained witnin acceptacle values

nrcushcut each fuel cycle.

OAVIS-3 ESSE, UNIT 1 3 3/a 1-1

REACTIVITY CONTROL SYSTEMS D

BASES' b'

3/4.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average teaperature less than 525'F.

This limitation is required to ensure 1) the moderator temperature coeffi-cient is within its analyzed temperature range, 2) the protective instrumentation is within its nomal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RT temperature.

NDT 2/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity contro-is available during each mode of facility operation. The components required to perfom this function include 1) barated water sources, 2) makeup or DHR pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE emergency busses.

With the RCS average temperature abava 200*F, a minimum of two separate and redundant baron injecti'.n systems are provided to ensure single functional capability in the eve at an assumed failure renders one of the systems inoperable. Allow dle aut-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of eitner system-is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.0% ak/k after xenon decay and cooldown to 200 F.

The maximum boration capability requirement occurs from full power equilibr1Tn xenon conditions and l

requires the equivalent of either 7373 gallons of 8742 ppm borated water from the boric acid storage tanks or 52,725 gallons of 1800 ppm borated water from the borated water storage tank.

The requirements for a minimum contained volume of 434,650 gallons of borated water in the borated water storage tank ensures the capa-bility for borating the RCS to the desired level.

The specified cuantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.

Therefore, the larger volume of barated water is specified.

With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the OAVIS-BESSE, UNIT 1 B 3/4 1-2 Amendment No..X, 3 3

l j.

i i

REACTIVITY CONTROL SYSTEMS i

BASES i

l

. BORATION SYSTEMS (Continued) 3/4.1.2 stable reactivity condition of the reactor and the additional restrictions i

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 SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200*F j

to 140*F. This condition requires either 8603 gallcns of 3742 ppm borated water from the boric acid storage system or 28,200 gallons of j

1300 ppm barated water from the borated water storage tank.

1 The contained water volume limits include allowance for water not available because of discharge line location and other physical charac-I teristics. The limits on contained water volus, and boron concentration j

ensure a pH value of between 7.0 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 st.ress corrosion cracking on raechanical systems and components.

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

2 3/J.1.3 MOVABLE CONTROL ASSEMBLIES i

The specifications of this section (1) ensure that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of a rod ejection accident. OPERABILITY of the control red position indicators 1

is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

1 The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. For example, misalignment of a safety or regulating rod requires.a restriction in THERMAL POWER. The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent 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, UNIT l1 B 3/4 1-3

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)

The maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analyses. Measurement with 7

> 525'F and with reactor coolant pumps operating ensures that lXE measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

The limitation on THERMAL POWER based on xenon reactivity is necessary to ensure that power peaking limits are not exceeded even with specified rod insertion limits satisfied.

The limitation on axial power shaping rod insertion is necessary to ensure that power peaking limits are not exceeded.

l DAVIS-BESSE, UNIT 1 B 3/4 1-4 Amendment No. 3 3 l

l l

a 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integ-rity durin Frep ency)g Condition I (Normal Operation) and II (Incidents of Moderate events by:

(a) maintaining the minimum DNBR in the core

> T.30 during normal operation and during short term transients (b) maintaining the peak linear power density < 18.4 kw/ft during normal operation, and (c) maintaining the peak p6er 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 3J-2 j

'and the insertion limit curves, Figures 3.1-2 and 3.1-3 are based on LOCA analyses which have defined the maximun linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200*F following a LOCA. Operation outside of the power-imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.

The power-imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-2 and 3.1-3 and if the steady-state limit QUADRANT POWER TILT exists. Additional conservatism is introducted by application of:

a.

Nuclear uncertainty factors.

b.

Thermal calibration uncertainty, c.

Fuel densification effects.

d.

Hot red manufacturing tolerance factors.

e.

Potential fuel rod bow effects.

The ACTION statements which permit limited variations from the basic requirsments are accompanied by additional restrictions which ensures that the original criteria are met.

The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

F Nuclear Heat Flux Hot Channel Factor, is defined as the maximum 9

local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions.

DAVIS-BESSE, UNIT 1 B 3/4 2-1 Amendment No. X,

33

J l

POWER DISTRIBUTION LIMITS

{

BASES f

F*aH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod on which i

minimum CNBR occurs to the average red power.

i i

It has been determined by extensive analysis of possible operating i

power shapes that the design limits on nuclear power pasking and on minimum DNER at full power are met, provided:

1 F

2 g 1 94; F

1 1

71 Power Peaking is not a directly observable quantity and therefore 4

limits have been established on the bar.tas of the AXIAL POWER IMBALANCE produced by the power peaking.

It has been detennined that the above hot 1

channel factor limits will be met provided the following conditions are maintained.

I 1

1.

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

I 2.

Regulating rod groucs are sequenced with overlapping groups as

]

required in Specification 3.1.3.5.

j 3.

The regulating rod insertion limits of Specification 3.1.3.6 i

are maintained.

i 4

AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER I

IMBALANCE is a measure of the difference in power between the i

top and bottom halves of the core. Calculations of core average axial peaking factors for many plants and measurements fran operating plants under a variety of operating conditions have been correlated with AXIAL PCWER IMBALANCE. The cor elation j

shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between the limits specified in Specification 3.2.1.

i The design limit power peaking factors are the most restrictive 4

calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod insertion and are the core CNBR design basis. Therefore, for operation at a fraction of RATED T'riERMAL POWER, the design limits are met. Whenusjnginceredetectorstomaktpowerdistribu-tion maps to determine Fg and F'g:

Meas a.

The measurement of total peaking factor, F

, shall be increased by 1.4 percent to account for makufacturing toler-ances and further increased by 7.5 per:ent to account for measurement error.

4' DAVIS-BESSE, UNIT 1 3 3/4 2-2 Anendment No.11 y

e-v v

s-g

1 1

l-POWER DISTRIBUTION LIMITS s

BASES b.

The measurement of enthalpy rise hot channel factor, F s

beincreasedby5percenttoaccountformeasurementeNo,r. hall For Condition II events, the core is protected from exceeding 20.4 kw/ft locally, and from going below a minimum DNBR of 1.30, by 4

automatic protection on power, AXIAL POWER IMBALANCE, pressure and temperature. Only conditions 1 through 3, above, are aandatory since 1

the AXIAL POWER IMBALANCE is an explicit input to the k ictor Protection

.l System.

The QUADRANT POWER TILT limit assures that the radial power distribu-1 tion satisfies the design values used in the power capability analysis.

l Radial power distribution measurements are made during startuo testing and periodically during power operation.

The QUADRANT POWER TILT limit at which corrective action is req'uired provides DNB and linear heat generation rate protection with x-y plane power tilts.

In the event the tilt is not corrected, the margin for uncertainty on F is reinstated by reducing the powsr by 2 percant for j

each percent of kilt in excess of the limit.

3/4.2.5 DNB PARAMETERS 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.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months speriodic surveillance of these parameters through instru-ment readout is Rufficient to ensure that the parameters are restored within-their limits following load changes and other expected transient l

operation. The 18 month periodic measurement of the RCS total flow rate using delta P instrumentation is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow I

such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

r DAVIS-BESSE, UNIT 1 B 3/4 2-3 Amendment No. 3 3

3/4.4 REACTOR COOLANT SYSTEM 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 1.30 during all normal operations l

and anticipated transients. With one reactor coolant pump not in operation in one loop, THEr' MAL POWER is restricted t,y the l

Nuclear Overpower Based on ACS Flcw and AXIAL POWER IMBALANCE and the

.4ucle.r Overpower Based on Pumo Monitors trip, ensuring that the ONBR will be maintained acave 1.30 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22"., whichever is more restrictive.

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STAN0M; however, single failure contiderations require placing a DHR loop into operation in tne snutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable outeof-service time.

3/4.J.2 and 3/4.4.3 SAFETY VALVES The pressuri:er 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 336,000 lbs per hour of saturated 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 shutdown.

In the event that no safety valves are OPERABLE, an operating CHR loop, con-nected to the RCS, provides overpressure relief capability and will prevent RCS overpressuri:ation.

Ouring operation, all pressuri:er code safety valves must be OPERABLE to prevent the RCS from being pressuri:ed above its safety limit of 2750 psig. The comoined relief capacity of all of these 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 accordance with the provisions of Section XI of the ASME Soiler and Pressure Code.

DAVIS-BESSE, UNIT 1 B 3/4 4-1 Amendment No. 3 3

REACTOR COOLANT SYSTEM BASES For a RPS high pressure trip setpoint r f 2300 psig, the maximun overshoot of the Reactor Coolant System pressure for a loss of feed-water (L0rd) event would be 2350 psig. Also, the LOFW is the maximum over-pressure anticipated transient.

The string inaccuracies and drift for the RPS high pressure trip are 15.29 psi, or 16 psi con-servatively. The maximum pressure peak for an anticipated transient is then 2366 psig.

The inaccuracies and drift for the string that controls the electromatic relief valve for the pressurizer are 16.75 psi, or 17 psi conservatively.

Included in this value is an inaccuracy of 4 psi and a drift of 7.5 psi for the transmitter. The 4 psi and 7.5 psi were combined by taking the square root of the sum of the squares, giving 8.5 psi. Subtracting 4 psi from 8.5 psi gives a value of 4.5 psi that is attributable to only the drift. The 8.5 psi was then added to inaccuracy and drift values for other components in the string '.o ob-tain i. total of 16.75 psi.

i The allowable value of >2385.5 psig is obtained by subtracting i

4.5 psi due to the drift from the trip setpoint of >2390 psig. The minimum lift pressure for the pressurizer electromatic relief valve is then (2400 17) psig = 2373 osig. Consequently, the resul-tant margin between the maximum pressure peak of 2366 psig and mini-mum lift pressure of 2373 psig for the pressurizer electromatic relief valve following an anticipated transient is 7 psi.

Thus, a 2300 psig RPS high pressure trip setpoint and the above values for the pressurizer electromatic relief valve will avoid actua-tion of the pressurizer electromatic relief valve during anticipated transients.

DAVIS-BESSE, UNIT 1 3 3/4 4-la Amendment No. 3 3

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capabin of accommodating 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 listt is based on providing enough water volume to prevent a reactor cralant system low pressure condition that would actuate the ReaGur Protection System or the Safety Featura Actuation System. The high level limit is based on providing enough steam volume to prevent a pressurizer niah 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 9fnimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM GENERATCJ 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 i

maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical 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 expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these chemistry limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking dwing plant operation would be limited by the limitttion of steam icaerator tube leakage between the primary coolant systaa and the secondary coolant system (primary-to-secondary leakage = 1 GPM).

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 8 3/4 4-2 w

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ML19340B075

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