ML20078Q104

Proposed Tech Specs Re Mods of Listed TSs Per Recommendation of GL 93-05

ML20078Q104
Person / Time
Site:	Davis Besse
Issue date:	12/06/1994
From:	CENTERIOR ENERGY
To:
Shared Package
ML20078Q102	List: ML20078Q098 ML20078Q104
References
GL-93-05, GL-93-5, NUDOCS 9412210157
Download: ML20078Q104 (23)

Text

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LAR'94-0007 REAC11VITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT - SAFETY ANO REGULATING ROD GROUPS -

LIMITING CONDITION FOR OPERATIONS

,r 3.1.3.1 All control (safety and regulating) rods shall' be OPERA 8LE and positioned within + 6.5% (indicated position) of their group average height.

1 APPLICABILITY: MODES l' and 2*.

ACTION:

a. With one or more control rods inoperable due to being imovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within one hour and be in at least HOT STAN08Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. .With more than one control rod inoperable or misaligned from its group average height by more than + 6.55 (indicated position), be in at least HOT STAND 8Y Elthin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With one control rod inoperab,le due to causes other than addressed by ACTION a, above, or misaligned from its group .

average height by more than + 6.55 (ind'.cated position), POWER OPERATION may continue proviiIed that within one hour either:

1. The control rod is restored to OPERABLE status within the above alignment requirements, or

2. The control rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a) An analysis of the potential ejected rod worth is perforined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the rod worth is deter-mined to be < 1.05 Ak at zero power and < 0.65%

ak at RATED THERMAL POWER for the remainifer of the I fuel cycle, and b) The SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and l

• See Special Test Exceptions 3.10.1 and 3.10.2.

Amendment No.178 DAVIS-BESSE, UNIT 1 3/4 1-19 INFORMATION ONLY 9412210157 941206 PDR ADOCK 05000346 P PDR

LAR 94-0007 P gs 11 REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - SAFETY AND REGULATING ROD GROUPS LIMTTING CONDITION FOR OPERATIONS ACTION: (Continued) c) A power distribution map is obtained from the incore detectors and F g and F"n are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and d) Either the THERMAL POWER level is reduced to s 60% of the THERMAL POWER allowable for the reactor coolant pump combination within one hour and within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Flux Trip Setpoint is reduced to s 70% of the THERMAL POWER allowable for the reactor coolant pump combination, or e) The remainder of the rods in the group with the inoperable rod are aligned to within i 6.5% of the inoperable rod within one hour while maintaining the position of the rods within the limits provided in the CORE OPERATING LIMITS REPORT; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The position of each control 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 asymmetric rod monitor is inoperable, then verify the individual rod position (s) of the rod (s) , with the inoperable asymmetric rod monitor at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least 2% in any one direction at least once every 92 3+ days.

DAVIS-BESSE, UNIT 1 3/4 1-20 Amendment No. 75%, f(g, f$2, 178

LAR 94-0007 P gs 12 REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - AXIAL POWER SHAPING ROD GROUP LIMITING CONDITION FOR OPERATION 3.1.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless fully withdrawn, and shall be positioned within i 6.5%

(indicated position) of their group average height.

APPLICABILITY: MODES 1* and 2*.

ACTION:

With a maximum of one APSR inoperable or misaligned from its group average height by more than i 6.5% (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 such 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 prevented while the rod remains inoperable or misaligned.

SURVEILLANCE REOUIREMENTS 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 r 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 asymmetric rod monitor is inoperable, then verify the individual rod position (s) of the rod (s), with the inoperable asymmetric rod monitor at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

.1.3.2.2 Uniccc al4 APSR arc fully .;ithdrawn, the APSR chall be determined to bc OPERAELE by moving the APSR rodo at 1:03: 2 *; at lecct oncc cvery 31 dayc.

• See Special Test Exceptions 3.10.1 and 3.10.2 DAVIS-BESSE, UNIT 1 3/4 1-21 Amendment No.162 l

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LAR 94-0007 Pcgs 13 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant Systes leakage shall be limited tot

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary. leakage through steam generators,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 10 GPH CONTROLLED LEAKAGE, and

f. 5 GPM leakage from any Reactor Coolant Systes Pressure Isolation Valve as specified in Table 3.4-2.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. Vith any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY vithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. Vith any Reactor Coolant System leakage greater than any one of'the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY vithin 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 55UTDOVN within the following 30,  !

hours except as permitted by paragraph c below.

c. In the event that integrity of any pressure isolation velive specified in Table 3.4-2 cannot be demonstrated, POVER OPERATION any continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a)

d. The provisions of Section 3.0.4 are not aoolicable .f.or entry into MODES 3 and 4 for the purpose of testing the isolation valves in l.

Table 3.4-2.

1 (8) Motor operated valves shall be placed in the closed position and power supplies deenergized.

!NFORMATION ONLY DAVIS-BESSE, UNIT 1 3/4 4-15 Olddt did. 4/10/M.

Amendment No. /l/36180

LAR 94-0007 Pcgs 14 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

b. Monitoring the containment sump level and flow indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Meacurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals to the makeup system when the Reactor Coolant System pressure is 2185 1 20 psig at least once per 31 days.

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant has been in COLD SHUTDOWN for 7"dayd 72 houra, or more, and if leakage testing has not"been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

4.4.6.2.3 Whenever the integrity of a pressure isolation valve listed in Table 3.4-2 cannot be demonstrated, determine and record the integrity of the high pressure flowpath on a daily basis. Integrity shall be determined by performing either a leakage test of the remaining pressure isolation valve, or a i combined leakage test of the remaining pressure isolation valve in a series with the closed motor operated containment isolation valve. In addition, record the position of the closed motor-operated containment isolation valve located in the high pressure piping on a daily basis.

Cidit 66 tid 04/2%/91 DAVIS-BESSE, UNIT 1 3/4 4-16 Amendment No.54, 115,180

LAR.94-0007 ,

Page 15 TABLE 3.A-2 REACTOR COOLANT SYSTEM PRESSURE.ISOLAT. ION, VALVES m r w auu 3 YALVE NUMBERS (b) MAXIMfl4 ALLOWABLE LEAXAGE (a)(c)

. SYSTEM CF-30 1 5 0 gps,

1. Decay Heat Removal Decay Heat Removal CF-31 1 5.0 gpa 2.

OH-76 1 5 0 gpa

3. Decay Heat Removal Cecay Heat Removal DN-77 1 5.0 gpm

4. ,

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l Notes:

(a) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

f

2. Leakage rates greater than 1.0 spa but less than or equal to 5.0 l gpa are considered acceptable if the latest measund rate has not - I exceeded the rate deterinined by the previous test by an amount that reduces the arrgin between measured leakage rate and the maximun l patinissible rate of 5.0 gpra by 50% or greater.

3. Leakage rates greater than 1.0 gpu out less than or equal to 5.0 gpa are considered unacceptable if the latest measured rate exceeded the rate determined by the previout' test by an amount that reduces the margin between measured leakage rate and the maximum periaissible rate of 5.0 gpa by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unaccept'a61e. ,

(b) Valves CF-30 and CF-31 will be tested with the Reactor Coolant system pressure >1200 psig. Valves OH-76 and DH-77 will be tested with norus) Core Flooding Tank pressure which is >S75 psig. Mini-mus differential test pressure across each valve shall not be less j than 150 psid, (c) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished i in accordance with approved procedures and suoported by computations showing that the method is capable of demonstrating valve compliance wIth the leakage criteria.

, t 3/4 e.-16a Order ctd. /20/81 DAVI$.BESSE, UNIT 1

LAR 94-0007 P;gs 16

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EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T . 2 280*F l

LIMITING CONDITION FOR OPERATION l

3.5.2 Two independent ECCS subsystems shall be OPERABLE with each j subsystem comprised of:

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

b. OneOPERABLElowpressureinjection(LPI)purg,

c. One OPERABLE decay heat cooler, and g '.

d An OPERA 8LE flow path capable of taking suction from the 81 # 1 borated water storage tank (8WST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: N00ES 1. 2 and 3.

ACTION:

- a. With one ECCS subsystem inoperab;s. restore the inoperable "

subsystem 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 HOT SHUTDOWN within the next I? hours,

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 Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation )

and the total accumulated actuation cycles to date.

SURVEILLANCE REOUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by vert'fying that each valve (manugl.

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 5-3 Ainendment No. M.182

LAR 94-0007 l Pagn 17  ;

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SURVElll ANCF RFOUIREMFNTS (continued) {

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b. At least once per 18 months or prior to operation after ECCS piping has been  :

drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points.

c. By a visual inspection which verifies that no loose debris (rags trash, clothing, etc.) is present in the containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditinns.

This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. Fofcontainment"entrieCduring!the"p#1od*when"CONTAINHENT"lNTEGRITYN' required.- by inspecting. Jet:least once dailyi:the afeast of containment affected by;the entry. and durir 'the final exitf(containmentLclosecutM~

gr the arcat ~affected Eith~ i e n c' n inst at^th6 spktfen 'bre icA ' '

cent 1 m nt entry ^ ^^ C0" "EMr turrcpirv Sc cetab'ished.

d. At least once per 18 months by:

1. Verifying that the interlocks:

a) Close DH-11 and DH-12 and deenergize the pressurizer heaters, if either DH-ll or DH-12 is open and a simulated reactor coolant system pressure which is greater than the trip setpoint (<438 psig) is applied. The interlock to close DH-ll and/or DH-12 is not required if the valve is closed and 480 V AC power is disconnected from its motor operators.

b) Prevent the opening of DH-11 cnd DH-12 when a simulated nr actual reactor coolant system pressure which is greater than the trip setpoint (<438 psig) is applied.

2. a) A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

b) Verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip. with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close in 575 seconds after the operator manually pushes the control switch toopentheContainmentEmergencySumpValveHV-DH9A(HV-DH9B)which shouad be verified to open in 575 seconds.

3. Verifying a total leak rate 5 20 gallons per hour for the LPl system at:

a) Normal operating pressure or hydrostatic test pressure of g 150 psig for those parts of the system downstream of the pump suction isolation valve.

and b) 2 45 psig for the piping from the containment emergency sump isolation valve to the pump suction isolation valve.

DAVIS-BESSE UNll 1 3/4 5-4 Amendment No. 3.25.28.40.77.135.182

1 ge 8 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

4. Verifying that a minimum of 72 cubic feet of trisodium l phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

5. Verify that a representative sample of TSP from a TSP

. storage basket has a density of 2 53 lbs/cu ft.

6. Verifying that when a representative sample of TSP from a TSP storage basket is submerged, without agitation, in at least one liter of 180 10*F borated water from the BWST, such that the resulting concentration of TSP is less than 0.84 grams per liter, the pH of the mixed solution is raised to 27 (measured at 77'F) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. e. At least once per 18 months, during shutdown, by

1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

2. Verifying that each HPI and LPI pump starts automatically upon receipt of a SFAS test signal.

f. By performing a vacuum leakage rate test of the watertight enclosure for valves DH-ll and DH-12 that assures the motor operators on valves DH-ll and DH-12 will not be flooded for at least 7 days following a LOCA:

1. At least once per 18 months.

2. After each opening of the watertight enclosure.

3. After any maintenance on or modification to the watertight enclosure which could affect its integrity.

g. By verifying the correct position of each mechanical position stop for valves DH-14A and DH-148.

1. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of the opening of the valves to their mechanical position stop or following completion of maintenance on the valve when the LPI system is required to be OPERABLE.

2. At least once per 18 months.

INFORMATION OND' DAVIS-BESSE, UNIT 1 3/4 5-5 Amendment No. J , Ap,191

LAR 94-0007 .

Pegs 19 '

.q -

j EMERGENCY CORE COOLING SYSTEMS

• i.

4 SURVEILLANCE REQUIREMENTS (Continued) i i

h. By performing a flow balance test, during shutdown, following completion of modifications to the HPI or LPI subsystems that alter the subsystem flow characteristics and verifying the '

following flow rates:

HPfSystem-SinglePump ,

Injection Leg 1-1 1 375 gpm at 400 psig*

Injection Leg 1-2 1 375 gpm at 400 psig*

Injection Leg 2-1 1 375 gpm at 400 psig*

Injection Leg 2-2 1 375 gpm at 400 psig*

LPI System ci,igle Pump Inj?ction Leg 1 1 2650 gpm at 100 psig** .

Injeci.on Leg 2 1 2650 gpm at 100 psig** . -

(.)

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• Reactor coolant pressure at the HPI nozzle in the reactor coolant pump discharge.

• • Reactor coolant pressure at the core flood nozzle on the reactor vessel.

INFORMATION ONLY

. I DAVIS-BESSE, UNIT 1 3/4 5-Sa Amendment No. 20

___ - - - - - - - - - _ - _ - - - - - - _ - - _ - - - _ - - - - - - - _ - - - - - - - - - -- - a

'LAR 94-0007-PIgn 20 .

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i INFORMATION ONLY.

CONTAINMENT SYSTEMS _

. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS

0NTAINMENT SPRAY SYSTEM _ '

I LIMITING CONDITION FOR OPERATION 3 .6.2.1 Two independent containment spray systems shall be OPERABLE  :

d ith each spray system capable of taking suction from the BWST on a  !

entainment spray actuation signal arid manualli transferring suction' l t o the containment emergency sump during the recirculation phase of .

3peration. .f APPLICABILITY: MODES 1, 2, 3 and 4. ,

ACTION:

dith one containment spray system inoperable, restore the inoperable s pray system 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 lea'st 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 />; restore the inoperable spray system i i to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDO] -

dithin the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,

2

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$URVEILLANCE REQUIREMENTS l f

4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

)

a. At least once per 31 days by verifying that each valve .

- (manual, power operated or automatic) in the flow path that j

' is not locked, sealed or otherwise secured in position, is in '

its correct position.  :

b. At least once per 18 months, during shutdown, by:

~

1. Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray test signal.

2. Verifying that each spray pump starts automatically on a SFAS. test signal.

INFORMATION ONLY .

3/4611 Amendment No. 36 DAVIS-BESSE, UNIT 1

'~-

7,_

.LAR 94-0007 Pcgs 21 ADDill0NAL CHANGES PREVIOUSLY  ;

J PROPOS!D BY LETTER

.CQRTAINMENT' SYSTEMS Serial No.J A 6 L Datelo*7 M SURVEILLANCE REOUIREMENTS (Continued)

c. At least once per 18 months by verifying a total leak rate s 20 -

gallons per hour for the system at:

1. Normal operating pressure or a hydrostatic test pressure

'of 2 150 psig for those parts of the system downstream of i_ the pump suction isolation valve, and 1

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2. 2 45 psig.for the' piping from the containment emergency i sump isolation valve to the pump suction isolation valve,

d. At least once per 10]G years by performing an air or smoke flow test through each" spray header and verifying each spray nozzle L is unobstructed.

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DAVIS-BESSE, UNIT 1 3/4 6-12

LAR 94-0007 P gs 22 l

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SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.4 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided:

a. Reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s), and

b. All axial power shaping rods are withdrawn to at least 35% (indicated position) and OPERABLE.

APPLICABILITY: MODE 2.

ACTION:

a. With any safety or regulating control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion or the axial power shaping rods not within their withdrawal limits, immediately initiate and continue boration at > 25 gpm of 7875 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all safety or regulating control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at > 25 gpm of 7875 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REOUIREMENTS 4.10.4.1 The position of each safety, regulating, and axial power shaping rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.4.2 Each safety or regulating control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within ?!daps?)4 hour-e prior to reducing the SHUTDOWN MARGIN to less than~~tihe ~

limits of Specification 3.1.1.1.

DAVIS-BESSE, UNIT 1 3/4 10-4 Amendment No. 191 i

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LAR 94-0007 l Page 23 Se<C1At 1<Sz e m ,110sS INFORMATION ONLY  :

l SHUT 00WN MARGIN SURVEILLANCE REQUIREMENTS (Continued) 4.10.4.3 The axial power shaping rods shall be demonstrated OPERABLE by i moving each axial power shaping rod > 6.5% (indicated position) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

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INFORMATION ONLY .

DAVIS-BESSE, UNIT 1 3/4 10-5

)

=A"" 1ivi14 <0mm SvSuMS INFORMATION ONLY l BASES 3/4.1.2 BORAT10N SYSTEMS (Contiente.dl d 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 to  ;

70*F. This condition requires either 700 gallons of 7875 ppm borated '

water from the BAAS or 3,000 gallons of 2100 ppm borated water from the

. BWST.

The bottom 4 inches of the BWST are not availcble, and the instrumen-tation is calibrated to reflect the available volume. All of the boric acid addition tank volume is available. The limits on water volume, and boron concentration 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 evclution 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 ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES 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 rod position indicators is i required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that 1 the original criteria are met. For example, misalignment of a safety or I regulating rod reqJires 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. 1 INFORMATION ONLY l

DAVIS-8 ESSE, UNIT 1 B 3/4 1-3 Amendment No. /123,191 l l

LAR 94-0007 Paga 25 l

REACT!*,'ITY CONTROL SYSTEMS _

! I BASES j

! ,' l I 3/4.1.3. MOVABLE CONTR01^ ASSEMBLIES (Continued) l

\

The maximum rod drop tu. pemitted is consistent with the assumed rod drop time used in the safety analyses. Measurement with T > 525'F and with

. reactor coolant pumps operating ensures that the meas 0EEd drop times will be representative of insertion times experienced during a reactor trip at j operating conditions.

{

Control rod positions and OPERABILITY of the rod position indicators are l' required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with frequent ie verifications required if an automatic monitoring channel is inoperable.

' These verification frequencies are adequate for assuring that the applicable

':LCO's are satisfied. A 1.5% group average position uncertainty is applied to

the roa index curves. .Therefore, the ' position indicators must be capable of ll supporting this accuracy. The Surveillance Requirement ensures this accuracy by keeping the RPI calibrated to a "known" position as indicated by the API.

Using the API as a "known" position is valid provided two consecutive reed switches are not inoperable. Having one entire string (i.e., every other reed switch) inoperable is acceptable.

A specific surveillance of the reed switches is not required because: i ,

,1) When one or more reed switch fails closed, a large API indication of

asymetry occurs. l l' 2) Two failed open reed switches in series result in a large indication

cf asymetry.

,, 3) Failsd open reed switches not in series (up to'every other switch) are

bounded by the analysis.

'l

[ Therefore, a reed switch condition not bounded by the analysis will be indicated by API system asymetry indications.

- Technical Specification 3.1.3.8 provides the ability to prevent excessive l power peaking by transient xenon at RATED THERMAL POWER. Operating restric-  !

tions resulting from transient. xenon power peaking, including xenon-free l startup, are inherently included in the limits of Sections 3.1.3.6 (Regulating l li Rod Ir.sertion Limits), 3.1.3.9 (Axial Power Shaping Rod Insertion Limits), and l

3.2.1 (Axial Power Imbalance) for transient peaking behavior bounded by the  ;

I  ; following factors. For the period of cycle operation where regulating rod  !

l groves 6 and 7 are allowed to be inserted at RATED THERMAL POWER, an 8% peaking l

FP. increase is applied F:r operation whereat or abova only 92% FP.

regulating rodAn 18%7increase group is allowed is applied below 92%

to be inserted j at RATED THERM,1 POWER, a 5% peaking increase is applied at or above 92% FP

- and a 13% incr.Me is applied below 92% FP. -

DAVIS-5 ESSE, UNIT 1 B 3/4 l-4 Amendment No. 45 i UMMATION,162ONIX

IAR 94-0007 Pcgs 26 l

REACTIVITY CONTROL SYSTEi45

{

BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)

If these values, checked every cycle, conservatively bound the peaking effects of all transier.t xenon, then the need for any hold at a power level cutoff below RATED THERMAL POWER is precluded. If not, either the power level at which the requirc:r.ents of Section 3.1.3.8 must be satisfied or the above-listed factors will be suitably adjusted to preserve the LOCA linear heat rate limits.

The limitation on axial power shaping rod insertion is necessary to ensure i

that power peaking limits are not exceeded.

l l

l l

i INFORMATION ONLY DAVIS-BESSE, UNIT 1 B 3/4 1-5 Amendment No. 33, 4 , 162

LAR 94-0007 Page 27

.aCroe C .L = 5'5 =

INFORMATION ONLY BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAGGE 3/4.4.6.1 LEAGGE DETECTION 56ird

• The RCS leakage detection systems required by this specification are i provided to detect and monitor leakage from the Reactor Coolant Pressure  !

soundary. These detection systasi. are consistent with the recommendation of Regulatory Guide 1.45. " Reactor Coolant Pressure Boundary Leakage Detectiois Systems.! May 1973. .

3/4.4.6.2 OPERATIONAL LEAKAGE PRES $URE 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 PRES $URE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SMUf00WN. .

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 l sufficiently low to ensure early, detection of additional leakage. t The total steam generator tube leakage Ifmit of 1 GPM for all steam generators ensures that the dosage contribution from tube leakage will be limited to a small fraction of 10 CFR Part 100 limits in the event of eitner a steam generator tube rupture or steam Ifne break. The 1 GPM limit is  !'

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

The 10 GPM IDENTIFIED LEAKAGE Ilmitation provides allowance for a limited amount of leakage fron known sources wuse presence will not interfere with the detection of UNIDENTIFIED LEAKAGE ty the leakage detection systems. ,

The CONTROLLEO LEAKAGE Itatt of 10 GPM restricts operation with a -

total RCS leakage from all RC pop seals in excess of 10 PM.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the prob- r ability of gross valve failure and consequent intersystem LOCA. Leakage ,

from the RCS Pressurc Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. t INFORMATION ONLY .

i DAY 15-l[55(, URIT 1 Amendment No.180 g 3f4 4,4

. i LAR %-0007 l

~""

j 3/4.5 EMERGENCY CORE COOLING SYSTEMS (EC q

l. BASES I

2/05.1 CORE FLOODING TANG The OPERABILITY of each core flooding tank ensures that a sufficient '

volume of borated water will be immediately forced into the reactor l vessel in the event the RCS pressure falls below the pressure of the  !

tanks. This initial surge of water into the vessel provides the initial

. cooling mechanism during large RCS pipe ruptures, t The limits on volume, boron concentratien and pressure ensure that the assumptions used for core flooding tank injection in the safety analysis are met.

The tank power operated isolation valves are considered to be

" operating bypasses in the context of IEEE Std. 279-1971 which requires ,

thatbypassesofaprotectivefunctionberemovedautomaticallywhenever '

permissive conditions are not met. In addition, as these tank isolation valves fail to meet single failure criteria, removal of power to the valves is required.  ;

The one hour limit for operation with a core flooding tank CFT) inoperable for reasons other than boron concentration not withi(n limits minimizes the time the plant is exposed to a possible LOCA event occurring with failure of a CFT, which may result in unacceptable peak cladding temperatures.

the condition mustWith be boron concentration corrected for one CFT within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72notnour within limits, limit was developed considering that the effects of reduced boron concentration on core subcriticality during reflood are minor. Boiling of the ECCS water in the core during reflood concentrates the boron in the saturated liquid that remains-in the core, in addition, the volume of the CFTs is still available for injection. Since the boron requirements are based on the average boron concentration of the total volume of both CFTs, the consequences are less severe than they would be if the contents of a CFT ,

were not available for injection.

The completion times to bring the  :

Limiting Condition for Operation (LCO) plant to a MODE in which thedoes not appl!

based on operating experience. The completion times allow plant L conditions to be changed in an orderly manner and without challenging  !

plant systems.  :

CFT boron concentration sampling within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after an 80 gallon ,

volume increase will identify whettier inleakage from the RCS has caused a reduction in boron concentration to below the required limit. It is not r necessary to verify boron concentration if the added water inventory is (

because the water contained  :

fromtheboratedwaterstoragetank(BWST)lonrequirements.

in the BWST is within CFT boron concentrat  !

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS l t

t The operability of two independent ECCS subsystems with RCS average .l temperature > 280 F ensures that sufficient emergency core cooling  ;

capability will be available in the event of a LOCA assuming the loss of 'i one subsystem through any single failure consideration. Either subsystem {

operating in conjunction with the core floodin tanks is capable of supplying sufficient core cooling to maintain p k claddin i i temperatureswithinacceptablelimitsforhf9 q A y 3 ty  !

ranging from the double ended break of the!Ikrge i Qlgg" h '

j downward. In addition, each ECCS subsystem prov es ong erm core cooling capability in the recirculation mode during the accident recovery period DAVIS-BESSE, UNIT 1 B 3/4 5-1 Amendment No. M ,191

LAR 94-0007

.Page 29 lNFORMATION ONL" EMERGENCY CORE COOLING SYSTEMS -

BASES l With the RCS temperature below 280*F, ona 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 l component ensures that, at a minimum the assumptions used in the safety analysesaremetandthatsubsystemOPERABILITYismaintained. The decay  !

heat removal system leak rate surveillance requirements assure that the i leakage rates assumed for tho system during the recirculation phase of the low pressure injection will not be exceeded. j The function of the trisodium phosphate dodecahydrate (TSP) contained in baskets in the containment normal sump is to neutralize the acidity of 1 the post-LOCA borated water mixture prior to establishing containment ,

emergency sump recirculation. The borated water storage tank (BWST) l borated water has a nominal pH value of approximately 5. Raising the -

borated water mixture to a pH value of 7 will ensure that chloride stress corrosion does net occur in austenitic stainless steels in the event that 1 chloride levels increase as a result of contamination on the surfaces of the reactor containment building. Also, a pH of 7 is assumed for the containment emergency sump for iodine retention and removal post-LOCA by the containment spray system.

The Surveillance Requirements JSR associated with TSP ensure that the minimum amount and density of 7SP)is stored in the baskets, and that the i TSP in the baskets is sufficient to provide adequate, post-LOCA, long-  !

term pH adjustment. ],

Surveillance requirements for throttle valve position stops and flow I balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure dro in the piping system to each injection point is necessary to: ) prevent total pump flow from exceeding runout conditions when( he system is in its minimum resistance configuration, (2) provide the proper flow split between injection with the assumptions used in the ECCS-LOCA analyses, and points in accordance (3) provide an acceptable level of total ECCS flow to all injection points equal to or i above that assumed in the ECCS-LOCA analyses. I l

Containment Emergency Sump Recirculation Valves DH-9A and DH-98 are de- 1 energized during MODES 1, 2, 3 and 4 to preclude postulated inadvertent l opening of the valves in the event of a Control Room fire which could j resultindrainingtheBoratedWaterStorageTanktothedontainment Emergency Sump and the loss of this water source for normal plant shutdown. Re-energization of DH-9A and DH-9B is permitted on an intermittent basis during MODES 1 2 3 and 4 under administrative controls. Stationproceduresidelitifytheprecautionswhichmustbe taken when re-energizing these valves under such controls.

outlet isolation valves DH-7A and DH-7B Borated Water Storage are de-energized Tank (BWST)2, during MODES 1, 3, and 4 to preclude postulated DAVIS-BESSE, UNIT 1 B 3/4 5-2 Amendment No. 20,191 123, 182 INFOR!WATION ONE

LAR 94-0007 Pagn 30 ,

Comi-1 SvsmS INFORMATION ONLY gm BASES i

3/4.6.1.4 INTERNAL PRESSURE [

The limitations on containment internal pressure ensure that 1) the f containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the containment peak pressure does not exceed the design pressure '

of 40 psig during LOCA conditions.

The maximum peak pressure obtained 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.

t 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that l 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- l ment steel vessel will be maintained comparable to the original design ,]

standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 38 psig i in the event of a LOCA. A visual inspection in conjunction with Type A j leakage tests is sufficient to demonstrate this capability. l 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The limitation on use of the Containment Purge and Exhaust System limits the time this system may be in operation with the reactor coolant system temperature above 200*F. This restriction minimizes the time -

that a direct open path would exist from the contairunent atmosphere to

+ht outside atmosphere ar.; consequently reduces the probability that an accident dose would exceed 10 CFR 100 guideline values in the event of a LOCA occurring coincident with purge system operation. The use of this system is therefore restricted to non-routine usage not to exceed 90 I hours in any consecutive 365 day period which is equivalent to approximately 1% of the total possible yearly unit operating time. l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM i

The OPERABILITY of the containment spray systen ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment .

DAVIS-BESSE UNIT 1 B 3/4 6-2 Anendment No. 135 IUORMATION ORY

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

LAR 94-0007 Page 31  ;

t - > - s,5 ,5 INFORMATION ONLY l BASES leakage rate are consistent with the assumptions used in the safety analyses.

The leak rate surveillance requirements assure that the leakage assumed for the system during the recirculation phase will not be exceeded.

Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and DH-78 are ide-energized during NODES 1, 2, 3, and 4 to preclude postulated inadvertent

closure of the valves in the event of a fire, which could result in a loss of

the availability of the BWST. Re-energization of valves DH-7A and DH-78 is ,

' permitted on an intermittent basis during MODES 1, 2, 3 and 4 under admints- l ltrative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.

4 Containment Emergency Sump Recirculation Valves DH-9A and DH-98 are de-ener-

'gized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent opening, of the valves in the event of a fire, which could result in draining the Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown. Re-energization of valves DH-9A and DH-98 is permitted on an intermittent basis during NODES 1, 2, 3, and 4 under administrative controls. Station procedures identify the precautions '

which must be taken when re-energizing these valves under such controls.

3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the . containment atmosphere or pressurization of the containment. Containment isolation within the required time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

Containment isolation valves and their required isolation times are addressed in the USAR. The opening of a closed inoperable containment isolation valve on an intermittent basis during plant operation is pennitted under administrative control. Operating procedures identify those valves which may be opened under administrative control as well as the safety precautions which must be taken when opening valves under such controls.

l INFORMATION ONLY l

DAVIS-BESSE. IINIT 1 B 3/4 6-3 Amendment No. JM, 717.182 I

LAR 94-0007 P gs 32 3/4.10 SPECIAL TEST EXCEPTIONS i

BASES 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their specified group heights and insertion limits and to be assigned to other than specified control rod groups, and permits AXIAL POWER IMBALANCE and QUADRANT POWER TILT limits to be exceeded during the performance of such PHYSICS TESTS as those required to 1) measure control rod worth, 2) detennine the reactor stability index and damping factor under xenon oscillation conditions and 3) calibrate AXIAL POWER IMBALANCE and QUADRANT POWER TILT instrumentation.

3/4.10.2 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER and is required to verify the fundamental nuclear characteristics of the reactor core and related instrumentation.

3/4.10.3 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under various flow conditions and is required in order to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 SHUT 00WN 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.

INFORMATION ONLY DAVIS-BESSE, UNIT 1 8 3/4 10-1