ML20138M128

Proposed Tech Specs 3.5.2 Re Emergency Core Cooling Systems & 4.5.2.f Re Surveillance Requirements

ML20138M128
Person / Time
Site:	Davis Besse
Issue date:	02/14/1997
From:	CENTERIOR ENERGY
To:
Shared Package
ML20138M120	List: ML20138M116 ML20138M125 ML20138M128
References
NUDOCS 9702250273
Download: ML20138M128 (10)

Text

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LAR.97-0006 ~'

Pcgs' 6 INFORMATION ONLY EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T . .llh 280*F l

LlHITlHG CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERA 3LE with each subsystem comprised of:

a.

One OPERABLE high pressure injection (HPI) pug, b.

Orie OPERABLE low pressure infection (LPI) pump,

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

An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and 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 inoperable, 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. '

In the event the ECCS is actuated and tnjects water into the Reactor Coolant System a Special Report spil te tretared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accuculated actuation cycles to,da,te.

SURYEILLANCE REOUTROiENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: -

a.

At least once per 31 days by verifying that each valve (manuti, power operated or automitic) in the flow path that is not'lotked, sealed or otherwise secured in position, is in its correct position.

DAYlS-BESSE. UNIT 1 3/4 S-3 /cendment No. II,182 '

M 4

9702250273 970214 PDR ADOCK 05000346 P

PDR

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INFORMATION ONUf  !

Revised by NRC Letter Dated )'

June 6. 1995 fSURVEILLANCEREQUIREMENTS(Continued) u At least once per 18 months, or prior to operation after ECCS piping has been

. b. I i drained by verifying that the ECCS piping is full of water by venting the ECCS l

pump casings and discharge piping high points. ** l

c. By a visual ins 4 clothing, etc.)pection which verifies that no loose debris (rags, trash.is present in

, containment emergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shall be performed:

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

2. For all areas of contairment affected by an entry, at least once daily while work is ongoing and again during the final exit after completion of work (containment closeout) when CONTAINMENT INTEGRITY is established.

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-11 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-11 and/or DH-12 is not required if  !

the valve is closed and 480 V AC power is disconnected from its motor operators.

i' b) Prevent the opening of DH-11 and DH-12 when a simulated or 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 veriffes ,

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.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 valvesener in <75 seconds after the operator manually pushes the control switch  :

to open the Containment Emergency Sump Valve HV-DH9A (HV-OH9B) which should be verified to open in <75 seconds.

3. Deleted The requirements of this surveillance may be deferred until the Tenth Refueling Outage l

for the ECCS flowpath which does not have manual high point venting capability.

DAVIS-BESSE. UNIT 1 3/4 5-4 -

Amendment No. 3.25.28.40.77.

135.182.195.196.208

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, $AR97-0006

. Paga 8 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

4. ' Verifying that a minimum of 290 cubic feet of trisodium phosphate j dodecahydrale (TSP) is contained within the TSP storage baskets.  !

i j 5. Deleted i

6. Deleted i

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

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

l L 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 i DH-11 and DH-12 that assures the motor operators on valves DH-11 and DH-12 will

! not be flooded for at least 7 days following a LOCA:

I 1. At least once per 18 months.

2. After each opening of the watertight enclosure.

' After any maintenance on or modification to the watertight enclosure which 3.

! could affect its integrity.

i The inspection nort on the watertight enclosure may be opened without reauiring nerformance of the vacuum leakage rate test. to nerform insnections. After use. the

! insocction nort must be verified as closed in its correct oosition. Provisions of TS 3.0.3 are not applicable during these insocctions.

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

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.

l DAVIS-BESSE, UNIT 1 3/45-5 Amendment No. 26,40,191,207,

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, T.

. L'AR 97-0006

~1 dORMAT10LONLY EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

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h. By perfcrming a flow balance test, durir.g shutdown, following l completion of modifications to the HPI or LPI subsystems that l alter the subsystem flow characteristics and verifying the .

(

following flow rates: .

HPI System - Single Pump -

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Injection Leg 1-1 > 375 gom.at 400 psig*

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

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

Irjection Leg 2-2 1 375 gpm at 400 psig* .

l/I System - Single Pump Infection Leg 1 . > 2650 gpm at 100 psig**

Injection Leg 2 E 2650 gpm at 100' psig**

.R eactor coolant pressure at the HPI nozzle in the reactor coolant pump discharge.

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

i DAVIS-BESSE. UNIT 1 3/4 5-Sa Araendr.ent No. 20

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j ,

LAR 97-0006 "

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Pac 3s lo

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3/4.5 ENERGENCY CORE COOLING EYSTEMS (ECCS)

as<s  !

{ 3/4.5.1 CORE FLOODING TANKS amMMAT!0N ONLY t

volume of borated. water will be immediately forced into the reacto I

i i vessel tanks. in the event the RCS pressure falls below the pressure of the j cooling mechanism during large RCS pipe ruptures.This initial surge of i The limits on volume boron concentration and pressure ensure that the

assumptions are met. us~ed for cor,e flooding tank injection in the safety analysis i

"o The tank power operated isolation valves are considered to be -

th t rating bypasses in the context of IEEE Std 279-1971 which requires

' perai passes of a protective function be remove.d automat call whenever lve conditions are not met. In addition as these tan

~ yalves valves fail to meet single failure criteria, rem; oval of powerto theisolation.

is required. -

i The one hou'r limit for operation with a core flood tank i inoperable for reasons other than boron concentration ith mits minimizes the time the plant is exposed to a possible LOCA event t

cladding temperatures. occurring with failure of a CFT, which may result in unaccepta i

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' must With bo'tedron withinconcentration 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The for one hour limit was CFTthenot condition within limits, deI that the effects of reduced ron concentration on core '

subcritica ity during reflood are minor. 8o111 of the ECCS wate in

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the core during reflood co that remains in the core. ncentrates the boron i the saturated 11 id available foi injection. In addition the vo' uma of the CFTs is ill Since the bor,on requiresents are based on the average boron concentration of the total volume of both CFTs the .

consequences.are were not availableless for severe than they would be if the conten,ts of a CFT injection.

LimitingThe completion Condition fortimes to bring the lant to a H00E in which the 0 erat based on operating exper ence. ion T eLC0 does nottimes completion 7.pplyallow are reasonableplant plant systems. conditions to be changed in an orderly manner and without challenging 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 whether inleakage from the RCS has caused a reduction in boron concentration to below the reautred limit. It is not -

necessary to verify boron concentration if the added water inventor *' is from in the the BWST bora'ted is within waterCFTstorage boron concentrattank (BWST) ion requirements becausethewatercontaIned 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS temperature t 280 F ensures that sufficient cyeseThe core coolin ~

operability of [ two indepe capability will be available.in the event of a L assumin the oss of one subsystem through any sin'gle failure consideration. Ei her subsystem operating in con unction with the core floodin tanks is ca able of supplying. suffic ent core cooling to maintain he peak clad ing-temperatures within acceptable limits for all postulated break sizes <

ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsy. stem provides long tenu core-cooling period capability in the recirculation mode during the accident recovery ,

i DAVIS-BESSE, UNIT I B 3/4 5-1 Amendment No. E0,191  !

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. EMERGENCY CORE COOLING SYSTEMS

. LAR 97-0006

• jlASES . Pags 11 V/ith the htCS temperature below 280 F, onc OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

The function of the trisodium phosphate dodecahydrate (TSP) contained in baskets located in the containment normal sump or on the 565' elevation of containment adjacent to the normal sump, is to neutralize the acidity of the post-LOCA borated water mixture during containment emergency sump recirculation. He borated water storage tank (BWST) 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 not occur in custenitic stainless steels in the event that chloride levels increase as a result of contamination on the surfaces of the ree.ctor 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.

'Ihe Surveillance Requirement (SR) associated with TSP ensures that the minimum required volume of TSP is stored in the baskets. The minimum required volume of TSP is the volume that will achieve a post-LOCA borated water mixture pH of 2 7.0, conservatively considering the maximum possible sump water volume and the maximum possible boron concentration. He amount to TSP required is based on the mass of TSP needed to achieve the required pH. However, a required volume is verified by the SR, rather than the mass, since it is not feasible to weigh the entire amount of TSP in containment. The minimum required volume is based on the manufactured density of TSP (53 lb/ft'). Since TSP can have a tendency to agglomerate from high humidity in the containment, the density may increase and the volume decrease during normal plant operation, however, solubility characteristics are not expected to change, Therefore, considering possible agglomeration and increase in density, verifying the minimum volume of TSP in containment is conservative with respect to ensuring the capability to achieve the i minimum required pH. The minimum required volume of TSP to meet all analytical requirements is 250 ft'. The surveillance requirement of 290 ft' includes 40 ft' of spare TSP as margin. Total basket  !

capacity is 325 ft'.  ;

j Decay Heat Removal System valves DH-11 and DH-12 are located in an area that would be flooded following a LOCA. These valyes are located in a watertight enclosure to ensure their operability up to seven days following a LOCA. Surveillance Requirements are provided to verify the acceptable leak i tightness of this enclosure. An inspection port is located on this watertight enclosure. which is typicalb used for performing inspections ins.ide the enclosure. During the vacumuleakage rate test. the inspection nort is in a closed position and subject to the_ test. This inspection port may be subsequently opened for use in viewing inside the enclosure. Opening this inspection nort.will not require nerformance of the Yacuum leakage rate test because of the desien of the closure filtmgdYhich will preclude leakage under LOCA conditions. when prop _cIlv mstalled. Proper installation include.Sindependent verification.

Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system toucach injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

DAVIS-BESSE, UNIT 1 B 3/4 5-2 Araendment No. 20,123,182,191,195,207,

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- LAR,91-ooo6

. Pega 12 ,s.

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re<ec<ucv coat ce0ti,<c Sv51E"S ESES (Continued)

INEMMWN M&

Containment Emergency Sump Recirculation Valves DH-9A and OH-98 are de-energized during MODES 1,. 2, 3 and 4 to preclude postulated inadvertent opening of the valves in the event of a Control Room 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 DH-9A and DH-98 is permitted on an intermittent basis during H0 DES 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.

Borated Water Storage Tank (BWST) outlet isolation valves OH-7A and DH-78 are de-energized during H0 DES 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 administrative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.

3 /4. 5.4 BORATED WATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on the BWST o

minimum v~ lume and boron concentration ensure that:

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1) sufficient water is available within containment to permit recirculation cooling flow to the core following manual switchover to the recirculation mode, and l

2) The reactor will remain at least 1% Ak/k subcritical in the cold condition at 70*F, xenon free, while only crediting 50% of the control rods' worth following mixing of the BWST and the RCS water volumes.

These assumptions ensure that the reactor remains subcritical in the cold condition following mixing of the BWST and the RCS water volumes.

With either the BWST boron concentration or BWST borated water temperature not within limits, the condition must be corrected in eight '

hours. The .eight hour limit to restore the temperature or boron concentration to within limits was developed considering the time required to change boron concentration or temperature and assuming that the contents of th'e BWST are still available for injection.

The bottom 4 inches of the BWST are not available, and the instrumentation is calibrated to reflect the available volume. The limits on water volume, and boron concentration ensure a'pH value of between 7.0 and 11.0 of the solution sprayed within the containment after a design basis accident. The pH band minimizes t.he evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.

DAVIS-BESSE, UNIT 1 B 3/4 5-2a Amendment No.191, 207

. Dock 3t Number 50-346 6 Licenzo Number NPF-3 Srrial Number 2448 Attachment 2

'Page 1 ENVIRONMENTAL ASSESSMENT Identification of Proposed Action This proposed action involves the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1, Operating License Number NPF-3, Appendix A, Technical Specifications (TS). A license amendment request is proposed by Toledo Edison to revise TS 3/4.5.2, " Emergency Core Cooling Systems, ECCS Subsystems - Tavg a 280*F" Surveillance Requirement 4.5.2.f.

Surveillance Requirement (SR) 4.5.2.f requires that the watertight enclosure which encloses the Decay Heat Removal (DHR) System isolation valves DH-11 and DH-12 be demonstrated operable by performing a vacuum leakage rate test of the enclosure. These valves are located in a common valve pit in the lower elevation of the containment vessel. The lower elevation in containment would be flooded following a Loss-of-Coolant Accident. Since the motor operators for these valves are not qualified for a submerged environment, the watertight enclosure is provided to cover the valve pit to ensure the valves will not be flooded for at least seven days following a LOCA. The watertight enclosure consists of the walls of the valve pit and large 1/4-inch deck plates attached to a steel frame, which cover the valve pit.

Surveillance Requirement 4.5.2.f requires that the vacuum leakage rate test be performed: 1) at least once per 18 months, 2) after each opening of the watertight enclosure, and 3) after any maintenance on or modification to the watertight enclosure which could affect its integrity. Technical Specification 3.0.3 was entered by the DBNPS on February 12, 1997 when it was determined that the Kamlok coupling inspection port on the enclosure had been opened after the vacuum leakage rate test had been performed. The inspection port had been opened in Mode 3 during heatup from the Tenth Refueling Outage to perform an American Society of Mechanical Engineer Boiler and Pressure Vessel (ASME)

Code visual Inservice Inspection (ISI) of portions of the decay heat piping located within the enclosure. The ISI is conducted with the piping at full pressure and temperature pursuant to the requirements of the ASME Code. Compliance with the ASME Code ISI requirements is mandated by TS 4.0.5.

Since the vacuum leakage rate test can only be performed in Mode 4, 5, or 6 due to the required testing conditions, opening the inspection port to perform the ISI per TS 4.0.5 in Mode 3 would then require the plant to be cooled down to at least Mode 5 in order to perform the vacuum leakage rate test.

On February 12, 1997, with the plant at 100% rated thermal power it was determined that the inspection port had been opened on May 24, 1996, during Mode 3 after the vacuum leakage rate test had been performed.

Technical Specification 3.0.3 was entered and a plant shutdown commenced.

This shutdown was terminated and the plant returned to 100% rated thermal power when the NRC staff granted a verbal enforcement discretion from performance of this SR 4.5.2.f on February 12.

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Docket Number 50-346

. Licensa Numb 3r NPF-3 Sarial Numb 3r 2448 Attachment 2

'Page I The DBNPS is proposing a follow-up license amendment to the granting of this enforcement discretion which would allow the inspection port on the watertight enclosure to be opened for inspections during Modes 1, 2 or 3 without requiring performance of the vacuum leakege rate test. After use, the inspection port would be required to be verified as closed in its correct position. Also, the provisions of TS 3.0.3 would not be applicable during these inspections in order not to require a plant shutdown due to the inspection port being open.

Need for the Proposed Action The changes proposed are needed to replace NRC staff's enforcement discretion verbally. granted on February 12, 1997 which allowed continued plant operation. This enforcement discretion remains in effect until the NRC approves the proposed license amendment to revise the TS requirements, or until the DBNPS enters Mode 4 in an outage of sufficient duration to perform the vacuum leakage rate test, which ever occurs first.

Environmental Impact of the Proposed Action As described in the Safety Assessment and Significant Hazards consideration for the proposed license amendment. Toledo Edison has determined that the ECCS structures, systems and components which could be affected by the proposed license amendment, will continue to be capable of performing their safety functions.

The proposed license amendment will reduce the potential for unduly -

requiring cooldown and heatup transitions of plant equipment, thus pre-serving the cycling margin between plant design and actual operating history.

The proposed license amendment involves a change to a requirement with respect to the use of plant components located within the restricted area ,

as defined in 10CFR Part 20. As discussed in the Safety Assessment and l Significant Hazards Consideration, this proposed license amendment does not involve a significant hazards consideration. The proposed change to allow continued plant operation does not alter source terms, containment isolation or allowable releases. In addition, the proposed change does j not involve an increase in the amounts, and no change in the types, of any radiological effluents that may be allowed to be released offsite. )

Furthermore, there is no increase in the individual or cumulative occupational radiation exposure.

With regard to potential non-radiological impacts, the proposed license amendment involves no increase in the amounts or change in types of any non-radiological effluents that may be released offsite, and has no other j environmental impact. j Based on the above, Toledo Edison concludes that there are no significant radiological or non-radiological environmental impacts associated with the proposed license amendment.

1

Dockst Number 50-346

, Licanza Number NPF-3 Sarial Numb 3r 2448 Attachment 2

• Page 3 ,

Alternatives to the Proposed Action Since Toledo Edison has concluded that the environmental effects of the proposed action are not significant, any alternatives will have only similar or creater environmental impacts. The principal alternative would be not to grant the license amendment. This would not reduce the environmental impacts attributable to the plant. Furthermore, it would force a cooldown of the plant after performing inspections utilizing the inspection port.

Alternative Use of Resources This action does not involve the use of resources not previously con-sidered in the Final Environmental Statement Related to the Operation of the Davis-Besse Nuclear Power Station, Unit Number 1 (NUREG 75/097).

Finding of No Significant Impact t

Toledo Edison has reviewed the proposed license amendment against the categorical exclusion criteria of 10CFR51.22(c)(9) for an environmental assessment. As demonstrated in the proposed license amendment's Safety Assessment and Significant Hazards Consideration the proposed changes do not involve a significant hazards consideration, do not increase the types or amounts of effluents that may be released offsite, and do not  ;

increase individual or cumulative occupational radiation exposures.

Accordingly, Toledo Edison finds that the proposed license amendment, if approved by the Nuclear Regulatory Commission, will have no significant impact on the environment and that no environmental assessment is required.

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