ML20117P531

Proposed Tech Specs,Supporting Conversion of DBNPS from 18 Month to 24 Month Fuel Cycle

ML20117P531
Person / Time
Site:	Davis Besse
Issue date:	09/17/1996
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20117P527	List: ML20117P526 ML20117P530 ML20117P531
References
NUDOCS 9609240189
Download: ML20117P531 (48)

Text

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4 LAR 95-0019

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INFORfAATION ONLY EMERGENCY CORE COOLING SYSTEMS l

ECCS SUBSYSTEMS - T . ,, A280*F l

LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. OneOPERABLEhighpressureinjection(HPI) pump,

b. One 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: MODES 1, 2 and 3.

l ACTION: .

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE <tatus 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 ne t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and l submitted to the Cocinission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to.da.te.

SURVElli.ANCE REOUIREMENTS i 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days b verifying that each valve (manu 1, power operated or automatic in the flow path,that is not lo ked, sealed or othentise secured in position is in its correct position.

DAVIS-BESSE UNIT 1 3/4 5-3 Anendment No. M 182 f

f O Vl R]ll L 1[F 9609240189 960917 PDR ADOCK 05000346 P PDR

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LAR 95-0019 Revised by NRC Letter Dated Page 7 1 June 6. 1995  !

SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS pi)ing is full of water by venting the ECCS pump casings and discharge piping hig1 pointsc**-

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

2. For all areas of containment 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 )~ayhj]3EllE@{Gj@@yfy 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 and DH-12 is o)en and a simulated reactor coolant system pressure which is greater tlan the trip setpoint (<438 psig) is a) plied. The interlock to close DH-11 and/or DH-12 is not required if t1e valve is closed and 480 V AC power is disconnected from its motor  ;

operators. '

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 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-DH7B) automatically close in s75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH9B) which should be verified to open in s75 seconds.

3. Deleted

• • The requirements n . + , ,,

of thi;

<m ., +sm ecce curveillance

<w,m,+ s .,s us % may,,m+

besdeferred

,..- - ,m.., u,ntil s u sthe mm uTenth Refueling

+ , ,-.,+ m,,

C5 lb. ib h0NbbbkNIh5bhbkhNkbbbI DAVIS-BESSE, UNIT 1 3/4 5-4 Amendment No. 3.25,28,40,77, 135,182.195,196,208.

is RM[QKeERY4?BRidNEGi#sl355E5)Kff@g?Ti5lF55?ff6515fMg3?5?2[d@d EIMsMMpgestingag gatamagrusgggeMraM M%

i LAR 95-0019 1

. >-8 EMERGENCY CORE C00LiflG SYSTEMS INFORMATION ONLY SURVEILLANCE REQUIREMENTS (Continued) l i

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4. Verifying that a minimum of 290 cubic feet of trisodium l

phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

5. Deleted l

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6. Deleted l

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-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:

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 mecha .ical 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.

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DAVIS-BESSE, UNIT 1 3/4 5-5 Amendment No. -2h 10,.IM , 207 '

INFORMATION ONLY

r' t LAR 95-0019 Page 9 INFORMATION ONLY EMERGENCY CORE COOLING SYSTEMS _

SURVEILLANCE REQUIREMENTS (Continued)

h. By performing a flow balance test, during shutdown, following I

completion of modifications to the HPI or LPI subsystems that J alter the subsystem flow characteristics and verifying the following flow rates:

HPI System - Single Pump -

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

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

Injection Leg 2-2 y,375 gpm at 400 psig*

LPI System - Single Pump Injection Leg 1 1 2650 gpm at 100 psig**

Inj,ection Leg 2 y_2650 gpm at 100 psig**

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• Reactor coolant pressure at the HPI nozzle in the reactor coolant~

' pump discharge.

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

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!NFORMATION ONLY DAVIS-BESSE, UNIT 1 3/4 5-Sa Ame nd.T.en t No. 20 i

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r LAR 95-0019 Page 10 CONTAINMENT SYSTEMS 3/4.6.5 SHIELD BUILDING EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent emergency ventilation systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one emergency ventilation system inoperable, restore the ino)erable system to OPERABLE status within 7 days or be in at least HOT STA4DBY within ,

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

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SURVEILLANCE REQUIREMENTS 4.6.5.1 Each emergency ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, '

from the control room, flow through the HEPA filters and charcoal l adsorbers and verifying that the system operates for at least 15 minutes.

b. At least once EsaiiREE0ElIIMITNTERVXU per 18 months or (1) after any i structural maiRfsii3HE6~~b~ri'DiF1 EPA ~ filter or charcoal adsorber i housings, or (2) following painting, fire or chemical release in any

! ventilation zone communicating with the system by:

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I DAVIS-BESSE, UNIT 1 3/4 6-28 Amendment No. 155

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. LAR 95-0019 Page 11 CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)

1. Verifying that the cleanup system satisfies the in-place penetra-tion and bypass leakage testing acceptance criteria of less than 1%

and uses the test procedure guidance in Regulatory Positions C.5.a. ,

C.S.c and C.S.d of Regulatory Guide 1.52. Revision 2. March 1978, and the system flow rate is 8.000 cfm i 10%; .

2. Verifying, within 31 c%ys after removal, that a laboratory analysis l of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2.

March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a* of Regulatory Guide 1.52. Revision 2. March 1978, for a methyliodide penetration of less than 1%: and

3. Verifying a system flow rate of 8.000 cfm i 10% during system ,

operation when tested in accordance with ANSI N510-1980.  ;

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, j within 31 days after removal, that a laboratory analysis of a represen-tative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2. March 1978 meets the laboratory testing criteria of Regulatory Position C.6.a* of Regulatory Guide 1.52. Revision 2. March 1978 for a methyliodide penetration of less.than 1%.

d. At least once )[c!3EFjg@G31Tj@)) per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 8.000 cfm 10%:

2. Verifying that the system starts automatically on any containment isolation test signal:

3. Verifying that the filter cooling bypass valves can be manually opened; and

• The test is performed in accordance with ASTM D 3803-1979 with the follow-ing conditions: 1) equilibrate for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> at 30 C/70% relative humidity (RH). 2) challenge for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 30 C/70% RH. 3) elution for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 30 C/70% RH.

DAVIS-BESSE UNIT 1 3/4 6-29 Amendment No. 43, 135. 155, 209

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. I IAR 95-0019 i l

Page 12 i i

CONTAINMENT SYSTEMS i

i SURVEILLANCE REQUIREMENTS (Continued) i I

4.

Verifying that each system produces a negative pressure of greater than or equal to 0.25 inches Water Gauge in the I l annulus within 4 seconds after the fan attains a flow rate of 8000 cfm i 10%. This test is to be perfonned with the flow path established prior to starting the EVS fan, and the other dampers associated with the negative pressure I

boundary closed.

e. After each complete or partial replacement of a HEPA filter i

' bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 8000 cfm 1 10%.

f.

After each complete or partial replacement of a charcoal l

adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 8000 cfm i 10%.

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I INFORMATION ONLY DAVIS-BESSE, UNIT 1 3/4 6-30 Amendment No. 3 J35,155

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• LAR 95-0019

• Page 13 I

1 PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independent control room emergency ventilation systems shall be OPERABLE. ,

APPLICABILITY: MODES 1. 2, 3 and 4.

ACTION:

1 With one control room emergency ventilation system ino)erable, restore the  !

ino)erable system to OPERABLE status within 7 days or )e in at least HOT STAiDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 l hours. )

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SURVEILLANCE REQUIREMENTS 4.7.6.1 Each control room emergency ventilation system shall be demonstrated OPERABLE:

a. At .least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is less than or equal to 110 F when the control room emergency ventilation system is operating,

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes.

c.

At least once structural Es?siiER^cT6fithTTIEPA^Tiiter or charcoal adsorberh3REFUEMi mai&f '

housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

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DAVIS-BESSE. UNIT 1 3/4 7-17 Amendment No. 135, 155  !

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LAR 95-0019 9 Page 14 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1. Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c and C.S.d of Regulatory Guide 1.52.

Revision 2 March 1978, and the system flow rate is 3300 cfm i10%;

2. Verifying, within 31 days after removal, that a laboratory analysis of a re]resentative carbon sample obtained in accordance

! with Regulatory )osition C.6.b of Regulatory Guide 1.52. Revision

2. March 1978, meets the laboratory testing criteria of

, Regulatory Position C.6.a* of Regulatory Guide 1.52, Revision 2.

March 1978, for a methyliodide penetration of less than 1%; and

3. Verifying a system flow rate of 3300 cfm *10% during system operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, >

within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a*

of. Regulatory Guide 1.52. Revision 2, March 1978, for a methyliodide penetration of less than 1%.

e. At least once INEi!REf;0gDMGjigN,JfByft!!j per 18 months by:

l 1. Verifying that the pressure drop across the combined HEPA filters

! and charcoal adsorber banks is less than 4.4 inches Water Gauge while operating the system at a flow rate of 3300 cfm 10%;

! 2. Verifying that the control room normal ventilation system is l

isolated by a SFAS test signal and a Station Vent Radiation High test signal; and l

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• The test is performed in accordance with ASTM D 3803-1979 with the following conditions: 1) equilibrate for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> at 30'C/70% relative humidity (RH), 2) challenge for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 30 C/70% RH, 3) elution for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 30 C/70% RH.

DAVIS-BESSE, UNIT 1 3/4 7-18 Amendment No. 134,135,155.209.

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l LAR 95-0019 Page 15 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

INFORMATION ONLY l

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3. Verifying that the makeup flow of the system is 300 cfm i 10% l when supplying the control room with outside air.

f. Af ter each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance witn ANSI N510-1980 for a D0P test aerosol while operating the system at a flow rate of 3300 cfm i 10%.

g. Af ter each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 3300 cfm i 10%.

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l lNFORMATION ONLY DAVIS-BESSE, UNIT 1 3/4 7-19 Amendment No.155 l

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LAR 95-0019 ~

• Page 16 e m 1 SxS m S INFORMATION ONI_Y 3/4.7.7 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.7 All safety-related snubbers shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located

. on systems required OPERABLE in those MODES).

ACTION:

a. With one or more snubbers inoperable: 1. within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber (s) to OPERABLE status, or 2. verify system operability nith the snubber (s) inoperable by engineering evaluation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; or 3. declare the supported subsystem inoperable and follow the appropriate ACTION statement for that -

. system.

and, for snubbers which have failed either the visual or functional test:

b. Perform an engineering evaluation within 90 days to determine if any safety-related system or component has l been adversely affected by the inoperability of the j snubber and if the snubber mode of failure has imparted a j significant effect or degradation on the supported '

component or system.1 The provisions of Technical Specification 3.0.4 are not applicable for the component or system.

SURVEILLANCE REQUIREMENTS  !

4.7.7 Each snubber 2 shall be demonstrated OPERABLE by the requirements of the following surveillance programs and pursuant to requirements of Specification 4.0.5.

4.7.7.1 Visual Inspection Program 1 Engineering evaluation is not required when a snubber is removed for surveillance testing provided it is returned to OPERABLE status within the requirements of ACTION statement a. l 2 Safety-related snubbers are listed in the latest revision of applicable surveillance test procedure (s). Snubbers may be added to, or removed from, safety-related systems and their assigned groups without a License Amendment.

DAVIS-BESSE, UNIT 1 3/4 7-20 Amendment No. W./YYY,161 l

INFORMATION ONI.Y

l l LAR 95-0019 Page 17 INFORMATION ONLY l PLANT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued)

a. General Requirements At least once per inspection interva', each group of snubbers in use in the Plant shall be visually inspected in accordance with Specification 4.7.7.1.b and 4.7.7.1.c.

Visual inspections may be performed with binoculars, or other visual support devices, for those snubbers that are difficult to access and where required to keep exposure as low as reasonably achievable. Response to failures shall be in accordance with Specific 6 tion 4.7.7.1.d.

b. Inspection Interval l The inspection interval may be applied on the basis of l 1

snubber groups. The snubber groups may be established based on physical characteristics and accessibility.

• Inaccessible snubbers are defined as those located: (a)inside j containment, (b) in high radiation exposure zones, or (c) in '

areas where accessibility is limited by physical constraints such as the need for scaffolding. l l

Each of the groups may be inspected independently according to the schedule determined by Table 4.7-5. The visual inspection interval for each snubber group shall be detennined based upon the criteria provided in Table 4.7-5, and the first inspection interval detennined using the criteria shall be based upon the previous inspection interval as established by the requirements in effect before amend-ment 161 .

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DAVIS-BESSE, UNIT 1 3/4 7-21 Amendment No. 9A,161 (next page is 3/4 7-21a)

INFORMATION ONLY l

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ma 95-e age 18 INFORMATION ONLY I TABLE 4.7-5  !

SNUBBER VISUAL INSPECTION INTERVAL l NUMBER OF UNACCEPTABLE SNUBBERS '

Population Column A l or Group Column B Column C Extended Interval Repeat Interval

] (Notes 1 and 2) (Notes 3 and 6) Reduced Interval (Notes 4 and 6) (Notes 5 and 6) 1 0 80 0 1 0 0 1M 0 2 1 4

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150

! 0 3 200 8 2 5 l 300 13 5 12 25 3

400 8 18 36 500 12 24 48

{ 750 20 40 78

1000 or greater 29 56 i 109 i

Note 1:

The next visual inspection interval for a snubber population or

! group size shall be determined based upon the previous inspection

'interval interval.and the number of unacceptable snubbers found during that Snubbers may be grouped, based upon their accessibility during power operation, as accessible or inaccessible. These 3

categories may be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use .that decision as the basis upon which to determine the next inspection interval for that group.

Note 2:

Interpolation between population or group sises _ and the number of unacceptable snubbers is permissible. Use next lover integer for the value of the limit for Columns A, B, or C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.

4 Note 3: If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the i

previous interval but not greater than 48 months.

Note 4: If the number of unacceptable snubbers is equal to er less than the number in Column B but greater than the number in Column A, the next inspection interval shall be the same as the previous interval.

INFORMATION ONLY DAVIS-BESSE, UNIT 1 3/4 7-21a Amendment No.161

LAR 95-0019

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INFORMATION ONLY Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the '

( number in Column B, the next interval shall be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B and C.

Note 6: The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months, with the exception that inspection of inaccessible snubbers may be deferred to the next shutdown when plant conditions allow five days for inspection.

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l DAVIS-BESSE, UNIT 1 3/4 7-21b (next page is 7-22) Amendment No.161 IMFORMATION ONLY

LAR 95-0019 Page 20 .

!~ INFORMATION ONLY . l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. Acceptance Criteria l A snubber shall be considered OPERABLE as a result of a visual inspection if: (1) there are no visible indica-tions of damage or inoperability, and (2) attachments to i

the foundation or supporting structure are secure, j

d. Response to Failures f or each snubber unit which does not meet the visual inspection acceptance criteria of Specification 4.7.7.1.c:

1. Determine the snubber OPERABLE by functionally testing the

[

snubber in the as-found condition per Specification l I 4.7.7.2, unless the (hydraulic) snubber was detennined .

l inoperable because the fluid port was found uncovered; and

2. Clearly establish and remedy the cause of the rejection l l

for that particular snubber and for other snubbers that may be generically susceptible; and

3. Classify the snubber as acceptable for the purpose of establishing the next visual inspection interval.

l. OR_

1. Perform the ACTION specified in 3.7.7a; and 1

2. Perform an engineering evaluation as specified in 3.7.7.b; and

3. Classify the snubber as unacceptable and establish the frequency of group inspection as described in l

Specification 4.7.7.1.b.

l e. Transient Event Inspection An inspection shall be performed of all hydraulic and mechanical snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as detennined from a review of operational data. A visual.

l inspection of the snubbers on these systems shall be perfonned within six months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber movement; or (2) evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.

DAVIS-BESSE, UNIT 1 3/4 7-22 Amendment No. f9,9f,JJJ,161 INFORMATION ONLY

IAR 95-0019 Page 21 I

I PLANT SYSTEMS INFORMATION ONLY i

SURVEILLANCE REQUIREMENTS (Continued) 5 l

4.7.7.2 Functional Test Program l

a. General Requirements At least once per inspection interval a representative sample of each group of snubber in use in the Plant shall be functionally tested in accordance with Specifications l 4.7.7.2.b and 4.7.7.2.c. Response to the failures shall be in accordance with Specification 4.7.7.2.d.

For all snubbers, functional testing shall consist of either bench testing or in place testing.

l b. Inspection Interval and Sample Criteria l

) The snubbers may be categorized into groups based on l physical characteristics and accessibility. Each group may be tested independently from the standpoint of l performing additional tests if failures are discovered.

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DAVIS-BESSE. Unit 1 3/4 7-22a Amendment No. JJJ./VM161 (Next page is 3/4 7-23

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. LAR 95.0019 Page 22 ,

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l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) ,

i The ins ection interval for functional testing shall be [@ 4 BE[0gljHEEjV_AL318 months.

1 Snubbers which are scheduled for removal for seal maintenance l l may be included in the test sample prior to any maintenance on j the snubber.

resentative sam le shall consist of at least 10 percent Therehdofftonexth$ghestinteger)ofeachgroupofsnubbers (round in use in the Plant. The selection process shall ensure that all snubbers, regardless of their accessibility classification, are functionally tested at least once every ten inspection intervals.

c. Acceotance Criteria For hydraulic snubbers (either inplace testing or bench test- l ing). the test shall verify that:

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1. Snubber piston will allow the hydraulic fluid to " bypass" Trom one side of the )iston to the other to assure unre-l strained action is aclieved within the specified range of l velocity or acceleration in both tension and compression.

2. When the snubber is subjected to a movement which creates a load condition that exceeds the specified range of velocity or acceleration, the hydraulic fluid is trapped in one end of the snubber causing suppression of that movement. ,

3. Snubber release rate or bleed rate, where required, occurs I in compression and tension.

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For mechanical snubber in place and bench testing, the test shall verify that:

1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

2. Activation (restraining action) is achieved in both tension and compression within the specified range.

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i DAVIS-BESSE. UNIT 1 3/4 7-23 Amendment No. 53, 94. 111

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l LAR 95-0019 Page 23

PLANT SYSTEMS INFORMATION ONLY i

SURVEILLANCE REQUIREMENTS (Continued)

d. Response to Failures For each inoperable snubber per Specification 4.7.7.2.c:

1. Perform the ACTIONS specified in 3.7.7a and 3.7.7b; af

2. Within the specified inspection interval, functionally test an additional sample of at least 10 percent of l the snubber units from the group that the inoperable

snubber unit is in.

The func:,Ional testing of an additional sample of at least 10 percent from the inoperable snubber's group is required for each snubber unit determined to be inoperable in subsequent functional tests, or until all snubbers in that group have been tested; and

3. The cause of snubber failure will be evaluated and, l l if caused by a manufacturing or design deficiency,  ;

all snubbers of the same or similar design subject to the same defect shall be functionally tested within 90 days from determining snubber inoperability. This i

testing requirement shall be independent of the requirements in 4.7.7.2.d(2) above.

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INF0ilMATION ONLY t DAVIS-BESSE, UNIT 1 3/4 7-24 Amendment No. 15,94 l (Tables 3.7-3 and 4.7-4 deleted. Next Page is 3/4 7-36).

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. LAR 95-0019 Page 26 REFUELING OPERATIONS i

STORAGE P0OL VENTILATION LIMITING CONDITION FOR OPERATION 3.9.12 Two inde)endent emergency ventilation systems servicing the storage pool area shall )e OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool.

ACTION:

a. With one emergency ventilation system servicing the storage pool area ino)erable, fuel movement within the storage pool or crane o)eration wit 1 loads over the storage pool may proceed provided the OPERABLE emergency ventilation system servicing the storage pool area is in o)eration and discharging through at least one train of HEPA filters and clarcoal adsorbers,

b. With no emergency ventilation system servicing the storage pool area OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one system is restored to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l

4.9.12.1 The above required emergency ventilation system servicing the storage pool area shall be demonstrated OPERABLE per the a nlicable S.urveil-lance Requirements of 4.6.5.1. and at least once eT6dRER!ilig pee 18 =nths by verifying that the emergency ventilariBKTistWirTbYvicinfthe storage pool area maintains the storage pool area at a negative pressure of

= 1/8 inches Water Gauge relative to the outside atmosphere during system operation.

4.9.12.2 The normal stora e 2001 ventilation system shall be demonstrated OPERABLE at least once per 18 =nths by verifying that the system fans stop au ay ampers automatically divert flow into the emergency ventilation system on a fuel storage area high radiation

. test signal.

DAVIS-BESSE. UNIT 1 3/4 9-12 Amendment No. 135. l

LAR 95-0019

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3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

INFORMATION ONLY BASES 3/4.5.1 CORE FLOODING TANKS The OPERABILITY of each core flooding tank ensures that a sufficient volume of borated. water will be immediately forced into the reactor vessel in the event the RCS pressure falls below the pressure of the tanks. This initial surge of water into the vessel provio6s the initial cooling mechanism during large RCS pipe ruptures.

, The limits on volume boron concer.tration and p. essure ensure that the assumptions used for cor,e flooding tank injection in the safety analysis

, are met. . i The tank power operated isolation valves are considered to be

" operating bypasses in the context of IEEE Std. 279-1971 which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition as these tank isolation valves fail to meet single failure criteria, rem; oval of power'to the yalves is required. ,

The one hour limit for operation with a core flooding tank (CFT) inoperable for reasons other than boron concentration not within 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. .

j the condition must With be.boron concentration corrected for oneThe within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. CFT 72not within hour limitlimits, 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 bor,on 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

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were not available for injection.

The completion times to bring the olant to a MODE in which the Limiting Condition for Operation (LCO) does not apply are reasonable based on operating experience. The completion times allow olant conditions to be changed in an orderly manner and without challenging plant systems.

CFT boron concentration samplina 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 whet 6er inleakage from the RCS has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is because the water contained from in thethe BWSTbora'ted water is within CFTstorage tank (BWST)lon r.equirements.

boron concentrat -

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The operability of two independent ECCS subsystems with RCS average l temperature g 280 F ensures that sufficient emeraency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any sin'gle failure consideration. Either subsystem operating in conjunction with the core flooding tanks is capable of supplying sufficient core cooling to maintain the peak cladding 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 subsystem provides long term core-cooling capability in the recirculation mode during the accident recovery period DAVIS-BESSE, UNIT 1 B 3/4 5-1 Ame tN AD 1

IA'l 95-0019 Page 26 EMERGENCY C0kt COOLING SYSTEMS BASES With the RCS temperature below 280'F, one OPERABLE ECCS subsystem is acceptable without single: failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem 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. The 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 austenitic stainless steels in the event that 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 Requirement (SR) associated with TSP ensures that the minimum required volume of TSP is stored in the baskets. Tha minimum required volume of TSP is the volume that will achieve a post-LOCA borated water mixture pH of h 7.0, conservatively considering the maximum possible sump water volume and the maximum possible boron concentration. The amount of TSP required is based on the mass of TSP needed to achieve the required W' 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 4

containment. The minimup required volume is based on the manufactured l 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 1 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 minimum required pH. The pinimum required volume of TSP to meet all analytical 3

requirements is 250 ft . The surveillance 3

requirement of 290 ft ingludes 40 ft of spare TSP as margin. Total basket capacity is 325 ft .

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 to each 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 l 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.

~l$E 0EIdATION 6fN' ^~" """"" ' " # "

LAR 95-0019 Page 27

< = = = m a< wou "c svs - s INFORMATION ONLY BASES (Continued)

Containment Emergency Sump Recirculation Valves DH-9A and DH-98 are de-energized during MODES I,. 2, 3 and 4 to preclude postulated inadvertent opening of the valves in the event of a Control Room fire, which could i 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 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.

Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and I'H-78 are de-energized during MODES 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 B0 RATED 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 minimum volume and boron concentration ensure that:

1) sufficient water is available within containment to permit

~ recirculation cooling flow to the core following manual switchover to the recircillation mode, and

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 4

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 the 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 the evolution of iodine and minimizes the effect of chloride and causti i o mechanical systems and components. f DAVIS-BESSE, UNIT I B 3/4 5-2a Amendment No. -l % 207

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LRR 95-0019 Page 28 INFORMATION ONUf CONTAINMENT SYSTEMS

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BASES .

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the Hydrogen Analyzers, Containment Hydrogen Dilution System, and Hydrogen Purge System ensures that this equipment will be available to maintain the maximum hydrogen concentration within the containment vessel at or below three volume percent following a LOCA.

The two redundant Hydrogen Analyzers deteratr.e the content of hydro-gen within the containment vessel. The Hydrogen Analyzers, although they have their OPERABILITY requirements in this Specification, are considered part of the post-accident monitoring instrumentation of Specificatt' on 3/4.3.3.6, Post-Accident Monitoring Instrumentation.

The Containment Hydrogen Dilution (CHD System consists of two full capacity, redundant, rotary, positive displ) a cement type blowers to supply air to the containment. The CHO System controls the hydrogen concentra-tion by the addition of air to the containment vessel, resulting in a pressurization of the containment and suppression of the hydrogen volume fraction.

The Containment Hydrogen Purge System Filter Unit functions in conjunction with the CHD System and is designed to release air from the containment atmosphere through a HEPA filter and charcoal filter prior to discharge to the station vent.

As a backup to the CHD System and the Containment Hydrogen Purge System, the capability to install an external hydrogen recombination system has been provided.

3/4.6.5 SHIELD BUILDING 3/4.6.5.1 EMERGENCY VENTILATION SYSTEM The OPERA 8ILITY of the emergency ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is neces-sary to meet the assumptions used in the safety. analyses and limit the site boundary radiation doses to within the' limits of 10 CFR 100 during LOCA conditions.

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INFORMATION ON0f DAVIS-BESSE. UNIT I B 3/4 6-4 . 67.183 ggN

LAR 95-0019 Page 29

,_g,,,,, INFORMATION ONLY BASES the flow path can be established. The ability for local, manual i

operation is demonstrated by verifying the presence of the handvheels for i all manual valves and the presence of either handwheels or available

' power supply for motor operated valves.

3/4.7.2 STEAM CENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures i that the pressure induced stresses in the steam generators do not j exceed the maximum allovsble fracture toughness stress limits. The j limitations of 110*F and 237 peig are based on a steam generator RT

of 40*F and are sufficient to prevent brittle fracture. NDT a

3/4.7.3 COMPONENT COOLING WATER SYSTEM 1

l The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related-equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

3/4.7.4 SERVICE WATER SiSTEM The OPERABILITY of:the service water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal'and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assum1tions 5 used in the safety analyses.

3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ~

ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.

• The limitations on minimum vater level and maximum temperature are based on providing a 30 day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants" March 1974.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYST M The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allovable temperature for continuous, duty rating for the equipment and instrumentation cooled by this system and 2) the control room vill remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A", l INFORMATION ONLY l DAVIS-BESSE, UNIT 1 B 3/4 7-4 Amendment No.103 I

LAR 95-0019

• Page 30 PLANT SYSTEMS INFORMATION ONU' I BASES PLANT SYSTEMS 3/4.7.7 SNUBBERS I

All safety-related snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a dynamic event.

Snubbers excluded from this inspection program are those installed on safety-related systems for loads other than dynamic or on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have n) adverse effect on any safety- l related system during a dynamic event.

Inoperable is defined as:

1. For visua1 test

a. The fluid no longer is supplied to the valve block, or

b. Mounting pins are disengaged from the snubber.

c. Attachment to foundation or supporting structure is not secure.

2. For. functional test

a. The snubber (excluding end anchors, i.e., pin-to-pin) does not meet specified test criteria.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval is determined by the number of inoperable snubbers found during an inspection, the total population or group size for each snubber type, and the previous inspection interval. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. Any inspection whose results require a shorter inspection l interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generi-cally susceptible, and verified by functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptibla snubbers are those which have checsame or similar design features directly l related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

DAVIS-BESSE, UNIT 1 B 3/4 7-5 Amendment No. H.III,161 INFORMATION ONLY

LAR 95-0019  ;

s Page 31 PLANT SYSTEMS BASES _

l When a snubber is found inoperable through a visual or functional test, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component  ;

or system has been adversely affected by the inoperability of the snubber. '

The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

res sam)le To of the installed provide snubbers of assurance will snubber be functionally tested functional reliability, REDEUNG a rep l JgERVAyat18monthintervals. Observed failures of these s mjife snubb~sFs sh Tl~fequire functional testing of additional units. When a snubber is found to be ino)erable due to failure to lock up or failure to move (i.e.. frozen in place), t1e cause will be evaluated for further action or testing.

In cases where the cause of failure has been identified, additional snubbers that have a high probability for the same type of failure or are being used in the same application that caused the failure shall be tested. This requirement increases the probability of locating inoperable snubbers.without  ;

testing 100% of the snubbers.

Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance programs.

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I DAVIS-BESSE. UNIT 1 B 3/4 7-Sa Amendment No. 94

LAR 95-0019 Page 32 REFUELING OPERATIONS i j ON ON&

BASES 3/4.9.12 STORAGE POOL VENTILATION The requirements on the emergency ventilation system servicing the storage pool area to be operating or OPERABLE ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.

3/4.9.13 SPENT FUEL p00L FUEL ASSEMBLY STORAGE The restrictions on the placement of fuel assemblies within the ' spent fuel pool, as dictated by Figure 3.9-1, ensure that the k-effective of the spent fuel pool will always remain less trian 0.95 assuming the pool to be flooded with non-borated water. The restrictions delineated in Figure 3.9-1 and the action statement are consistent with the criticality safety analysis performed for the spent fuel pool.

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1 lNFORMATION ONLY DAVIS-BESSE, UNIT 1 B 3/4 9-3 Amendment No. # 1, 135 l

• LAR 95-0019 Enclosure 1

( Page 1 Summary of Licensino Basis, Surveillance Data, and Maintenance Record Reviews for Surveillance Requirement 4.5.2.d.4 l

1. A. Technical Specification (TS) 3/4.5.2, " Emergency Core Cooling Systems, ECCS Subsystems - T a 280*F," Surveillance Requirement (SR) 4.5.2.d.4 *#9 1

l Note: Other TS 3/4.5.2 SRs that are affected by the 24 month cycle conversion will be addressed by separate License Amendment Requests.

B. Systems or Components:

None

, This surveillance measures the amount of trisodium phosphate

! dodecahydate (TSP) that is inside the containment building.

C. Updated Safety Analysis Report (USAR) Sections: 6.2.2.2.2,

" Containment Spray System," 6.3.3.2, " Additional Considerations for ECCS Performance," and 9.3.3.2, " Post-LOCA Sump pH-Control" l

l l 2. Licensing Basis Review:

1 A. SR 4.5.2.d.4 requires that at least once per 18 months, the l operability of the ECCS subsystems be demonstrated by verifying that a minimum volume of trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets which are located

! inside the containment building.

l l It is proposed that in SR 4.5.2.d the words "at least once per 18 l months" be replaced with "at least once each REFUELING INTERVAL."

l A separate License Amendment Request, (LAR 95-0018; DBNPS letter Serial Number 2342) proposes that " REFUELING INTERVAL" be defined as "a period of time s 720 days." Technical Specification 4.0.2 would continue to apply which would allow increasing the new surveillance interval on a non-routine basis from 24 months to 30 months.

B. This surveillance verifies that there is a sufficient volume l

of TSP contained in baskets inside the containment building.

TSP is used to adjust the pH of the borated water mixture that would be inside of containment in the event of a Loss of Coolant Accident (LOCA). The pH is adjusted to a value of 7 with TSP in i order to reduce the possibility of chloride stress corrosion of austenitic stainless steels. 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 use of TSP is discussed in USAR Sections 6.3.3.2 and 9.3.3.2.

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• LAR 95-0019 Enclosure 1 Page 2 l

l l 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 the required msss of TSP would remain available and the solubility characteristics would not be affected. Extending the surveillance from 18 months to 24 months will not affect the mass of TSP available.

i l The method of adjusting the post-LOCA borated water mixture using TSP is completely passive. Extending the surveillance will not affect any accident initiators, or affect the consequences of an accident.

C. The current surveillance interval of 18 months was based on the Technical Specifications issued with the original operating license by the NRC dated April 22, 1977. The proposed changes follow the guidance of Generic Letter 91-04, " Changes in Techni-cal Specification Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991.

D. As a result of the above review, it is concluded that the licensing basis for the TSP baskets inside the containment will not be invalidated by increasing the surveillance interval for SR 4.5.2.d.4 from 18 months to 24 months and by continuing to allow the application of TS 4.0.2 on a non-routine basis.

E.

References:

1. Davis-Besse Nuclear Power Station (DBNPS) Unit No. 1, Operating License NPF-3, Appendix A, Technical Specifica-tions, through Amendment 211.

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11. Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991. )

i lii. Davis-Besse Technical Specifications issued with the original operating license by the NRC dated April 22, 1977.

iv. USAR Section 6.2.2.2.2, " Containment Spray System," through Revision 19.

v. USAR Section 6.3.3.2, " Additional Considerations for ECCS Performance," through Revision 19.

vi. USAR Section 9.3.3.2, " Post-LOCA Sump pH-Control," through Revision 19.

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LAR 95-0019 Enclosure 1 Page 3

3. Surveillance Data Review:

A. The 18 month TS surveillance test data for SR 4.5.2.d.4 was reviewed for the period of the Fifth Refueling Outage (5RFO)

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i?icu3h 9RFO. This time period was selected because it reflects the major plant improvements after June 1985, and covers five refueling outages and four operating cycles of test results.

B. The test results indicate no failures or significant degradation.

The surveillance test data under review indicate that the minimum TSP volume acceptance criterion was always met.

C. Based on a review of the 18 month surveillance test results data, no additional actions are necessary or recommended to support this increase in the present surveillance interval.

D. Based on a review of the 18 month surveillance test results data, and given that the method of adjusting the post-LOCA borated water mixture using TSP is c/mpletely passive, it is concluded that it is acceptable to increase the surveillance interval for SR 4.5.2.d.4 from 18 to 24 months and that there is no adverse effect on nuclear safety. Furthermore, it remains acceptable to allow the continued application of TS 4.0.2 on a non-routine basis.

E.

References:

DBNPS Procedures DB-ST-5051.06 and DB-CH-03003, j

" Trisodium Phosphate Chemical Check."

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4. Maintenance Records Review: l l

The method of adjusting the post-LOCA borated water mixture using TSP l contained in baskets is completely passive. The TSP baskets require no maintenance, therefore, a maintenance records review is not applicable.

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I LAR 95-0019 Enclosure 2 Page 1 Summary of Licensing Basis, Surveillance Data, and Maintenance Record Reviews for Surveillance Reauirements 4.6.5.1.b, 4.6.5.1.d. 4.9.12.1, and 4.9.12.2 i

1. A. Technical Specification (TS) 3/4.6.5.1, " Containment Systems -

Shield Building, Emergency Ventilation System," Surveillance i Requirements (SR) 4.6.5.1.b and 4.6.5.1.d, and TS 3/4.9.12, j

" Refueling Operations, Storage Pool Ventilation," SR 4.9.12.1 and 4.9.12.2.

B. Systems or Components:

Shield Building EVS Storage Pool EVS C. Shield Building EVS: Updated Safety Analysis Report (USAR)

Sections: 6.2.3, " Containment Vessel Air Purification and Cleanup Systems."

Storage Pool EVS: USAR Sections: 9.4.2.2, " Fuel Handling Area,"

and 3D.2.13, " Safety Guide 13 ' Fuel Storage Facility Design Basis'(March 1971)."

2. Licensing Basis Review:

A. SR 4.6.5.1.b requires that at least once per 18 months, or, (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, demonstrate operability by: (1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria, (2) Verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample meets the laboratory testing' criteria, (3) Verifying the required system flowrate.

SR 4.6.5.1.d requires that at least once per 18 months, demonstrate operability by: (1) Verifying the maximum pressure drop across the combined HEPA filters and charcoal adsorber banks, (2) Verifying that the system starts automatically on any containment isolation test signal, (3) Verifying that the filter e voling bypass valves can be manually opened, and (4) Verifying that each system produces the required negative pressure in the annulus within 4 seconds after the fan achieves the required flow rate. This test is to be performed with the flow path established prior to starting the EVS fan, and the other dampers associated with the negative pressure boundary closed.

SR 4.9.12.1 requires that two independent emergency ventilation systems servicing the storage pool area shall be demonstrated operable at least once per 18 months per the applicable require-ments of SR 4.6.5.1, and by verifying that the emergency ventila-tion system servicing the storage pool area maintains the storage

o LAR 95-0019 Enclosure 2 Page 2 pool area at the required negative pressure relative to the outside atmosphere during system operation.

SR 4.9.12.2 requires that the normal storage pool ventilation system be demonstrated operable at least once per 18 months by verifying that the system fans stop automatically and that the dampers automatically divert flow into the emergency ventilation system on a fuel storage area high radiation test signal.

It is proposed that in SR 4.6.5.1.b, 4.6.5.1.d, 4.9.12.1, and 4.9.12.2, the words "at least once per 18 months" be replaced with "at least once each REFUELING INTERVAL." A separate License Amendment Request (LAR 95-0018; DBNPS letter Serial Number 2342) proposes that " REFUELING INTERVAL" be defined as "a period of time s 730 days." Technical Specification 4.0.2 would continue to apply, which would allow increasing the new surveillance interval on a non-routine basis from 24 months to 30 months.

B. The function of the Shield Building Emergency Ventilation System (EVS) is to collect and process potential leakage from the Containment Vessel to minimize environmental activity levels resulting from all sources of containment leakage following a loss-of-coolant accident. Each of the two redundant subsystems is provided with prefilters, high efficiency particulate air (HEPA) filters and charcoal adsorbers to remove airborne parti-cles and methyl iodide as well as elemental iodine contaminants i resulting from a LOCA. The charcoal adsorber section of the EVS is comprised of two beds in series, each two inches thick. The plant design and safety analysis is based on one two inch thick bed. The additional installed two inch thick bed provides redundancy. No credit is taken for'the second bed in the LOCA safety analysis.

The EVS exhausts air from the containment annulus and penetration rooms to provide a negative pressure in the annulus. Makeup air is introduced into these areas by infiltration through the mechanical penetration and ECCS Rooms, Shield Building leakage, and by potential Containment Vessel leakage.

Following a loss-of-coolant accident, a safety features actuation signal starts the Emergency Ventilation System fans and opens the dampers located in the penetration rooms' outlet ductwork. The safety features actuation signal also closes all containment isolation valves and purge system valves. The purge system fans, if running, are shut down automatically.

The Storage Pool EVS utilizes the Emergency Ventilation System (EVS) fans and filter banks. System operation is discuss 1d ir.

USAR Section 9.4.2.2. Ventilation is switched to EVS nn a hi<*

radiation signal in the fuel handling storage pool area. In the event of a fuel-handling accident, the radiation detectors would send signals to the essential solenoid valves and to the EVS fans. The essential solenoid valves would cause dampers on the

I i

r

. LAR 95-0019 Enclosure 2 Page 3 suction and discharge side of fuel-handling area supply and exhaust fans to close (this would cause supply and exhaust fans to stop) and the dampers in the bypass ducting to open. This action will ensure that the fuel-handling area supply and exhaust ducting are isolated and the EVS fans are started automatically to pull a negative pressure in the fuel handling area. This ensures that accident doses at the site boundary will be well below the 10CFR100 guidelines.

l The EVS is not an initiator, nor a contributor to the initiation i

of an accident described in the USAR. The EVS is designed such that a single failure of an active component in either subsystem does not affect the functional capability of the other subsys-tems. The EVS is designed as Seismic Class I.

l C. The surveillance intervals of 18 months for the EVS were originally based on the guidance of NUREG-0103, Revision 0, June 1, 1976, Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors. License Amendment 155 amended this l specification to incorporate the guidance of Regulatory Guide l 1.52 Revision 2, " Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," dated March, 1978. Regulatory Guide 1.52 provides l guidance for an 18 month testing interval, therefore the proposed l 24 month testing interval would be an exception to Regulatory Guide 1.52. This exception would be added to the DBNPS USAR to reflect a revised licensing basis.

D. As a result of the above review, it is concluded that the licensing basis of the Shield Building EVS Filters and the Storage Pool EVS will be appropriately revised to reflect an increase in the surveillance intervals for SR 4.6.5.1.b, 4.6.5.1.d, 4.9.12.1, and 4.9.12.2 from 18 months to 24 months and the continued allowance of the application of TS 4.0.2 on a non-routine basis.

As previously noted, extension of the surveillance interval to 24 months will be an exception to the recommendation of Regulatory Guide 1.52, Revision 2. Upon approval of this license amendment request, USAR Section 6.2.3.1 will be revised to incorporate a discussion of the exception.  ;

E.

References:

+

1. Davis-Besse Nuclear Power Station (DBNPS) Unit No. 1, Operating License NPF-3, Appendix A, Technical Specifications, through Amendment 211.

ii. Generic Letter 91-04, " Changes in Technical l

Specification Surveillance Intervals to Accommodate a i 24-Month Fuel Cycle," dated April 2, 1991.

I

+

LAR 95-0019 i Enclosure 2 Page 4 l 111. " Standard Technical Specifications for Babcock and  :

Wilcox Pressurized Water Reactors," NUREG-0103, '

Revision 0, dated June 1, 1976. I l

iv. USAR Section 6.2.3, " Containment Vessel Air Purification and Cleanup Systems," through Revision 19.

i

v. USAR Section 9.4.2.2, " Fuel-Handling Area," through Revision 19.

vi. USAR Section 3D.2.13, SAFETY GUIDE 13 " Fuel Storage Facility Design Basis."

vii. Generic Letter 83-13, " Clarification of Surveillance Requirements for HEPA Filters and Charcoal Adsorber Units in Standard Technical Specifications on ESF Cleanup Systems," dated March 2, 1983.

viii. Regulatory Guide 1.52, Revision 2, March 1978, " Design, Testing, and Maintenance Criteria for Post Accidert Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."

ix. Amendment 155, Davis-Besse Nuclear Power Station, dated April 23, 1991, Toledo Edison Log 3452.

3. Surveillance Data Review:

A. The applicable 18 month TS Surveillance Test results data for the EVS were reviewed for the period of the Fifth Refueling Outage (5RFO) through 9RFO. These tests are normally performed non-outage. This time period was selected because it reflects the major plant improvements after June 1985, and covers five j refueling outages and four operating cycles of test results. '

B. During the review period there were three occurrences where individual charcoal bed samples failed to meet SR 4.6.5.1.b requirements for methyl iodide penetration. Each time the spent charcoal bed was replaced. However, due to the additional series charcoal bed, the overall total system iodide penetration always complied with SR 4.6.5.1.b. The charcoal bed efficiency also always exceeded the value assumed in the accident analysis.

Therefore, there were no failures which would have prevented EVS from performing its function.

C. Based on a review of the 18 month surveillance test results data, no additional actions are required to support the increase in the present surveillance interval.

D. The failures noted in Section B above were due to normal use and aging. Experience shows that a charcoal bed will last approxi-mately six years. Plant Engineering is trending the laboratory

I

• LAR 95-0019 )

Enclosure 2 Page 5 l results of the charcoal bed testing in order to predict the need for charcoal bed replacement. The purpose of the trending pro-gram is to provide for charcoal bed replacement prior to loss of train operability. Extending the surveillance interval will not affect the use or aging of the charcoal beds. Under a 24 month fuel cycle, this will provide for two surveillance test periods before reaching six years, which provides ample time to identify charcoal beds approaching a need for replacement.

In accordance with TS SR 4.6.5.1.a, flow is passed through the HEPA filters and charcoal adsorbers on a staggered test basis, at least once per 31 days to demonstrate system operability. Therefore the components will not be experiencing a longer period of disuse or a ,

longer period between operation after extending the surveillance interval to 24 months.

Based on the above information, extending the surveillance inter-val from 18 months to 24 months will not increase the failure rate or introduce new failure modes. Therefore, this increase is acceptable. Furthermore, it remains acceptable to allow the continued application of TS 4.0.2 on a non-routine basis.

E.

References:

1. DBNPS Procedure DB-SS-03252, " Emergency Ventilation System Train 1 18-Month or Special Test."

11. PSNPS Procedure DB-SS-03253, " Emergency Ventilation System Train 2 18-Month or Special Test."

iii. DBNPS Procedure DB-SS-03254, " Emergency Ventilation System Train 1 18-Month SFAS Drawdown Test."

iv. DBNPS Procedure DB-SS-03255, " Emergency Ventilation System 1 Train 2 18-Month SFAS Drawdown-Test."

v. DBNPS Procedure DB-SS-03708, " Spent Fuel Pool Ventilation System 18-Month Test."

4. Maintenance Records Reviews A. Maintenance records were reviewed for the period of SRFO through 9RFO. This time period was selected because it reflects the major plant improvements after June 1985, and covers five refueling outages and four operating cycles of maintenance activities. ,

l B. Below is a listing of equipment failures during the period of SRFO through 9RFO.

Mechanical Penetration Room Purge Solenoid valves SV5009, SV5016, and SV5021 were found-failed, not allowing their valves CV5009, CV5016, and CV5021 respectively, to open. The SVs were replaced.  !

• LAR 95-0019 Enclosure 2 Page 6 The SV failures were generic problems solved by component replacement. These valves are normally closed, which is their required accident position. Therefore, the failures did not affect system operability.

Fuel Handling Area Supply Fan Supply Damper HAS404A failed to stroke fully closed. The system did, however, pass the surveillance test and operability was not affected. The damper ,

was adjusted, and no further action was required.

EVS Fan C30-1 had wear on the shaft due to loose sheave bolting.

The fan was repaired. Operability of EVS was not affected.  !

Containment Annulus Pressure Differential Controller PDC 5014 failed to calibrate. The controller was replaced. Although the controller could not be calibrated it was still able to control the contain-ment annulus pressure greater than a negative .25 inches water gauge (w.g.), therefore operability was not affected.

EVS Fan C30-1 had an overload light illuminate during testing.

It was determined that the fan breaker had a loose connection, which was subsequently repaired. However, during the test the system continued to perform its function.

ECCS Room Isolation Damper HA5439 failed due to a failed SFAS relay which was then replaced The failure did not affect system operability.

It is important to note that there were no repetitive type failures noted in the system review.

C. Based on the maintenance history review, no additional actions are required to extend the TS surveillance interval from 18 months to 24 months.

i D. Based on the historical good performance of these components, the lack of repetitive type failures, the low potential for significant increases in failure rates of these components under a longer test interval and the introduction of no new failure modes, it is concluded that it is acceptable to increase the surveillance interval of SR 4.6.5.1.b, SR 4.6.5.1.d, SR 4.9.12.1, and SR 4.9.12.2 from 18 to 24 months and that there will be no adverse effect on safety. The charcoal beds have been determined to last approximately six years before needing replacement. Under a 24 month fuel cycle, this will provide for two surveillance test periods before reaching six years, which provides ample time to identify charcoal beds approaching a need for replacement.

Furthermore, it is acceptable to allow the continued application of TS 4.0.2 on a non-routine basis.

E.

References:

1. DBNPS Maintenance Work Order Records ii. DBNPS Preventive Maintenance Records

I l

f a LAR 95-0019 Enclosure 3 Page 1 l Summary of Licensing Basis, Surveillance Data, and Maintenance Record Reviews for Surveillance Requirement 4.7.6.1.c and 4.7.6.1.e I

1. A. Technical Specification (TS) 3/4.7.6.1, " Plant Systems - Control Room Emergency Ventilation System," Surveillance Requirements (SR) 4.7.6.1.c and 4.7.6.1.e B. Systems or Components:

Control Room EVS l C. Updated Scfety Analysis Report (USAR) Sections: 9.4.1 " Control Room-Air Conditioning, Heating, Cooling and Ventilating System,"

12.2.1 " Ventilation-Design Objectives," and 12.2.2.4 " Ventilation-Control Room

2. Licensing Basis Review:

A. SR 4.7.6.1.c, requires that at least once per 18 months, or, (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, demonstrate operability by: (1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria; (2) verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample meets the laboratory testing criteria; and (3) Verifying the required system flowrate.

SR 4.7.6.1.e, requires that at least once per 18 months, demon-strate operability by: (1) Verifying the maximum pressure drop across the combined HEPA filters and charcoal adsorber banks; (2) ,

verifying that the control room normal ventilation system is i isolated by a SFAS test signal, and a Station Vent Radiation High test signal; and (3) Verifying that the makeup flow of the system l 1s 300 cfm i 10% when supplying the control room with outside i air.

It is proposed that in SR 4.7.6.1.c and SR 4.7.6.1.e, the words "at least once per 18 months" be replaced with "at least once each REFUELING INTERVAL." A separate License Amendment Request (LAR 95-0018; DBNPS letter Serial Number 2342) proposes that j

" REFUELING INTERVAL" be defined as "a period of time 5 730 days."

Technical Specification 4.0.2 would continue to apply which would allow increasing the new surveillance interval on a non-routine basis from 24 months to 30 months.

B. The Control Room Emergency Ventilation System (CREVS) consists of two 100 percent-capacity (3300 cfm) redundant fan-filter assem-blies. Each filter system includes a roughing filter, high-efficiency filter, and charcoal adsorber. A cooling coil and i

I .

• LAR 95-0019 Enclosure 3 Page 2 water-cooled condensing unit are provided for each system to provide suitable temperature conditions in the control room for operating personnel and safety-related control equipment. Two 100 percent-capacity redundant air-cooled condensing units are l

provided as a backup to the water-cooled condensing units. On high refrigerant head pressure, the Service Water Valve closes l and the refrigerant solenoid valves align the air-cooled conden-i sing unit automatically. The CREVS is designed to maintain a temperature of 95*F or below in the control room and shift manager's office. The CREVS is capable of maintaining 0.15 2

inches w.g.' positive pressure in the control room with a 1.5 ft leakage area from the control room with air at 75'F. The CREVS is

]

designated as Seismic Class I.

Area radiation monitors are provided for the control room which j continuously give the background radiation level. In case of any l abnormal increase in the background level, the operator can  ;

manually isolate the normal ventilation system and start the Control Room Emergency Ventilation System. The CREVS functions to mitigate the consequences of certain design basis accidents by pressurizing the control room and providing filtered recirculated air to control room personnel.

Extending the surveillance intervals will not affect any accident initiators, or affect the consequences of an accident.  ;

C. The surveillance interval of 18 months for Control Room EVS was originally based on the guidance of NUREG-0103, Revision 0, ,

June 1, 1976, Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors. As described in License Amendment 155, the current surveillance requirements of the Control Room EVS filters are in accordance with the recommenda- i tions of Generic Letter 83-13 " Clarification of Surveillance Requirements for HEPA Filters and Charcoal Adsorber Units in Standard Technical Specifications on ESF Cleanup Systems." The' ;

surveillances performed conform to the guidance of Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Post l Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear i Power Plants," Revision 2, dated March 1978. Regulatory Guide I 1.52, provides guidance for an 18 month testing interval, there-fore the proposed 24 month testing interval would be an exception to Regulatory Guide 1.52. This exception would be added to the DBNPS USAR to reflect a revised licensing bases.

l D. As a result of the above review, it is concluded that the licen-l sing basis of the Control Room EVS will be revised to reflect an 1 l

increase'in the surveillance intervals for SR 4.7.6.1.c and 4.7.6.1.e from 18 months to 24 months and by the continued allowance of the application of TS 4.0.2 on a non-routine basis.

l l

f

• LAR 95-0019 -

Enclosure 3 Page 3  :

As previously noted, extension of the surveillance interval to 24 months will be an exception to the recommendation of Regulatory Guide 1.52, Revision 2. Upon approval of this license amendment request, USAR Section 6.2.3.1 will be revised to incorporate a I discussion of the exception.

E.

References:

1

1. Davis-Besse Nuclear Power Station (DBNPS) Unit No. 1, Operating License NPF-3, Appendix A, Technical Specifica-tions, through Amendment 211.

ii. Generic Letter 91-04, " Changes in Technical Specification l Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991.

iii. " Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," NUREG-0103, Revision 0, dated June 1, 1976.

1 USAR 6.2.3, " Containment Vessel Air Purification and iv.

Cleanup Systems," through Revision 19.

I

v. USAR Section 9.4.1, " Control Room-Air Conditioning, Heat-ing, Cooling and Ventilating System," through Revision 19.

vi. USAR Section 12.2.1, " Ventilation-Design Objectives,"

through Revision 19.

vii. USAR Section 12.2.2.4, " Ventilation-Control Room," through Revision 19.

viii. USAR Chapter 15, " Accident Analysis," through Revision 19.

ix. Generic Letter 83-13, " Clarification of Surveillance Requirements for HEPA Filters and Charcoal Adsorber Units in Standard Technical Specifications on ESF Cleanup Systems," dated March 2, 1983.

x. Davis-Besse Nuclear Power Station License Amendment 155, dated April 23, 1991, Toledo Edison Log Number 3452.

xi. Regulatory Guide 1.52, Revision 2, March 1978. " Design,

. Testing, and Maintenance Criteria for Post Accident

! Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled l

Nuclear Power Plants."

3. Surveillance Data Review:

i f A. The applicable 18 month TS Surveillance Test results data for the j CREVS was reviewed for the period of the Fifth Refueling Outage

'(5RFO) through 9RFO. In addition, the test results for Control

. . , - . ~ . ~ . . - - - _ - -- - -. =. _ _ _ _ __

i lc LAR 95-0019-l Enclosure 3 I l Page 4 l  !

t l Room Isolation Damper Closure Times (performed for SFAS Surveillance 4.3.2.1.3) and the test results for maintaining a j positive pressure in the Control Room (performed per USAR 9.4.1) were also reviewed. The time period between 5RFO and 9RFO was  !

i selected because it reflects the major plant improvements after June 1985, and covers five refueling outages and four operating cycles of test results.

B. The test results indicate the following failures occurred over i the period of SRFO through 9RFO for these components.

, During 5RFO, Control Room Isolation Damper HV5311E closed slowly l (7.32 seconds) under an SFAS signal. Although this closure time '

met SR 4.7.6.1.e.2 (which requires closure with no specific time constraints), the time, when combined with the SFAS Logic Actua-tion Time, would have exceeded the required response time of 10 seconds required by TS Table 3.3-5 item 2.c, 4.c and 6.c (TS ,*

3/4.3.2.1). The root cause was determined to be excessive dirt.

The damper was cleaned and lubricated and was successfully retested.

During SRFO, the CREVS Train 1 charcoal sampled per SR 4.7.6.1.c.2 ,

did not meet the laboratory testing criteria of Regulatory Guide t 1.52 for a methyl iodide penetration of less than 1%. The char-coal was therefore replaced with new charcoal. ,

l' i During 5RFO, the CREVS Train 2 charcoal sampled per SR 4.7.6.1.c.2 l did not meet the laboratory testing criteria of Regulatory Guide i

! 1.52 for a methyl iodide penetration of less than 1%. The char-  !

coal was replaced with new charcoal.

i Prior to and during 5RFO there was no trending program in place to predict the need.for charcoal bed replacement. Plant Engineer-ing is now trending the laboratory results of these tests to i predict the need for charcoal-replacement prior to end of life. ['

l Laboratory results of samples since then show a slow degradation

(as expected), but no further failures. Based on the trending of l charcoal sample results, CREVS Train 2 charcoal was replaced in l 9RFO, after a 0.892% methyl iodide penetration test result.

C. Based on a review of the 18 month surveillance test results data, l no additional actions are necessary or recommended to support the l increase in the present surveillance interval.

D. The failures noted in Section B, above, were due to normal use and aging. Experience shows that a charcoal bed will last approximately six years. Plant Engineering is trending the laboratory results of the charcoal bed testing in order to predict the need for charcoal bed replacement. All components (including the charcoal filters and isolation dampers) have demonstrated the ability to perform their function over two 18 month cycles without failure; therefore, a 30 month interval is acceptable.

O a LAR 95-0019 Enclosure 3 Page 4 Room Isolation Damper Closure Times (performed for SFAS Surveillance 4.3.2.1.3) and the test results for maintaining a positive pressure in the Control Room (performed per USAR 9.4.1) were also reviewed. The time period between 5RFO and 9RFO was selected because it reflects the major plant improvements after June 1985, and covers five refueling outages and four operating cycles of test results.

B. The test results indicate the following failures occurred over the period of SRFO through 9RFO for these components.

During SRFO, Control Room Isolation Damper HV5311E closed slowly j (7.32 seconds) under an SFAS signal. Although this closure time j met SR 4.7.6.1.e.2 (which requires closure with no specific time i constraints), the time, when combined with the SFAS Logic Actua-tion Time, would have exceeded the required response time of 10 seconds required by TS Table 3.3-5 item 2.c, 4.c and 6.c (TS 3/4.3.2.1). The root cause was determined to be excessive dirt.

The damper was cleaned and lubricated and was successfully retested.

During 5RFO, the CREVS Train 1 charcoal sampled per SR 4.7.6.1.c.2 did not meet the laboratory testing criteria of Regulatory Guide 1.52 for a methyl iodide penetration of less than 1%. The char-coal was therefore replaced with new charcoal.

During SRFO, the CREVS Train 2 charcoal sampled per SR 4.7.6.1.c.2 did not meet the laboratory testing criteria of Regulatory Guide 1.52 for a methyl iodide penetration of less than 1%. The char-coal was replaced with new charcoal.

Prior to and during SRFO there was no trending program in place to predict the need for charcoal bed replacement. Plant Engineer-ing is now trending the laboratory results of these tests to predict the need for charcoal replacement prior to end of life.

Laboratory results of samples since then show a slow degradation (as expected), but no further failures. Based on the trending of charcoal sample results, CREVS Train 2 charcoal was replaced in 9RFO, after a 0.892% methyl iodide penetration test result.

C. Based on a review of the 18 month surveillance test results data, no additional actions are necessary or recommended to support the increase in the present surveillance interval.

D. The failures noted in Section B, above, were due to normal use and aging. Experience shows that a charcoal bed will last ,

approximately six years. Plant Engineering is trending the laboratory results of the charcoal bed testing in order to predict the need for charcoal bed replacement. All components (including the charcoal filters and isolation dampers) have demonstrated the ability to perform their function over two 18 month cycles without failure; therefore, a 30 month interval is acceptable.

O LAR 95-0019 Enclosure 3 Page 5 I

t Components of the Control Room EVS are exercised by other, more frequent (non-refueling) surveillances, such that the components will not be experiencing a longer period of disuse or a longer period between operation. Therefore, there is a low potential l for any significant increases in failure rates of the components l under a 24 month operating cycle. L l The increase in surveillance interval will not introduce any new  !

failure modes.

( It is therefore concluded that it is acceptable to increase the l surveillance interval for SR 4.7.6.1.c and SR 4.7.6.1.e from 18 to 24 months and that there is no adverse effect on nuclear ,

safety. Furthermore, it remains acceptable to allow the contin- I ued application of TS 4.0.2 on a non-routine basis. i

(. E.

References:

1. DBNPS Procedure DB-SS-03710, " Control Room Emergency Ventilation System Train 1 18-Month Surveillance Test."

11. DBNPS Procedure DB-SS-03711, " Control Room Emergency Ventilation System Train 2 18-Month Surveillance Test."

t lii. DBNPS Procedure DB-SS-03145, " Control. Room Emergency Ventilation System 18-Month or Special Test Train 1."

iv. DBNPS Procedure DB-SS-03146, " Control Room Emergency Ventilation System 18-Month or Special Test Train 2."

v. Regulatory Guide 1.52, Revision 2, March 1978, " Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-cooled Nuclear Power Plants."

vi. USAR Section 9.4.1, " Control Room-Air Conditioning, Heat-ing, Cooling and Ventilating System," through Revision 19.

4. Maintenance Records Reviews A. The 18 month CREVS maintenance records were revieved for the l period of the Fifth Refueling Outage (5RFO) through 9RFO. This j time period was selected because it reflects the major plant  ;

improvements after June 1985, and covers five refueling outages l_ and four operating cycles of maintenance activities. 1 l I l The Preventive Maintenance (PM) activities included the follow- l i ing: Inspecting the Control Room Isolation Damper actuators for  ;

. air leaks, inspecting damper blades and seals, repair / replacement '

l of blades / seals as necessary, seal lubrication, and stroking ,

dampers every refueling outage. These PMs help to ensure the l I

i l e l c LAR 95-0019 l

Enclosure 3 i Page 6 dampers are in good condition to be able to isolate the Control Room when needed. The performance of the actual Surveillance Tests (ST) verify the operability of these dampers.

B. Only one " failure" occurred during this time period. During 5RFO, Control Room Isolation Damper HV5311E closed slowly (7.32 seconds) under an SFAS signal. The cause of the excessively slow  ;

movement was dirt. The damper and its pivot points were cleaned, lubricated, and stroked four times, then that portion of the Surveillance Test was re-run, resulting in a satisfactory closure time.

C. Based on a review of the 18-month maintenance records, no addi-tional actions are necessary or recommended to support the  !

increase in the present TS surveillance interval.  !

D. Based on review of the 18-month maintenance data, particularly failure mechanisms, the CREVS components have demonstrated the ability to perform their function over two 18-month cycles without failure.

The CREVS components are exercised by other, more frequent (non-refueling) surveillances, such that the components will not be experiencing a longer period of disuse or a longer period between operation. Therefore, there is a low potential for any significant increases in failure rates of the component with the longer surveillance test interval.

The increase in surveillance interval will not introduce any new failure modes.

Based on the historical good performance of these components, the low potential for significant increases in failure rates, and the introduction of no new failure modes, it is concluded that it is acceptable to increase the surveillance interval from 18 to 24 months with no adverse effect on safety. Furthermore, it is acceptable to allow the continued use of TS 4.0.2 on a non-routine basis.

E.

References:

The following maintenance records were used:

1. DBNPS Maintenance Work Order Records.

ii. DBNPS Preventive Maintenance Records.

l l

f I

i

I l e LAR 95-0019 l Enclosure 4 Page 1 I

Summary of Licensing Basis, Surveillance Data, and Maintenance Record Reviews for Surveillance Requirement 4.7.7.2.b

1. A. Technical Specification (TS) 3/4,7.7, " Plant Systems - Snubbers,"

Surveillance Requirement (SR) 4.7.7.2.b B. Systems or Components Snubbers C. Updated Safety Analysis Report (USAR) Sections: Section 3,

" Design Criteria - Structures, Components, Equipment and Systems."

2. Licensing Basis Review:

A. SR 4.7.7.2.b requires that the inspection interval for snubber functional testing shall be 18 months.

It is proposed that in SR 4.7.7.2.b the phrase "18 months," be replaced with "each REFUELING INTERVAL." It is proposed that in TS Bases 3/4.7.7 the phrase "at 18 month intervals," be replaced with "each REFUELING INTERVAL." A separate License Amendment Request, (LAR 95-0018; DBNPS letter Serial Number 2342) proposes that " REFUELING INTERVAL" be defined as "a period of time s 730 days." Technical Specification 4.0.2 would continue to apply which would allow increasing the new surveillance interval on a non-routina basis from 24 months to 30 months.

B. The operability of the snubbers ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a dynamic event.

Extending the surveillance interval from 18 months to 24 months will not affect any accident initiators, or affect the consequences of an accident.

C. The current surveillance interval of 18 months was originally based on the guidance of NUREG-0103, Revision 0, June 1, 1976, Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors. As discussed in the NRC's Safety l Evaluation related to Amendment Nos. 201 and 204 (dated May 16, 1995) to the Peach Bottom Atomic Power Station, Unit 2 and 3 Operating Licenses DPR-44 and -56, respectively, the original interval of 18 months was selected mainly to accommodate the need to test snubbers which were inaccessible during operation. The sample size of 10 percent was arbitrarily selected which, when l applied with the 18-month interval, could lead to a test period

of 15 years for 100 percent of the snubbers. This was perceived l

to be a conservative test period for snubbers. Under a 24-month

! interval, the test period could be 20 years, however, the NRC

r l

• LAR 95-0019 Enclosure 4 Page 2 j l

Staff determined that changing the inspection cycle to 24 months would not significantly impact the functional testing program to confirm the operability of the snubber population. The proposed changes follow the guidance of Generic Letter 91-04, " Changes in Technical Specification Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991.

D. As a result of the above review, it is concluded that the licen-sing basis of the plant systems snubbers will not be invalidated  !

by increasing the surveillance intervals for SR 4.7.7.2.b from 18 i months to 24 months and by continuing to allow the application of TS 4.0.2 on a non-routine basis. Changing the inspection cycle to 24 months (with a maximum of 30 months) will not reduce the ability of the functional testing program to confirm the operability of the snubber population.

E.

References:

1

1. Davis-Besse Nuclear Power Station (DBNPS) Unit No. 1, l Operating License NPF-3, Appendix A, Technical Specifica-tions, through Amendment 211.

ii. Generic Letter 91-04, " Changes in Technical Specification .

Surveillance Intervals to Accommodate a 24-Month Fuel l Cycle," dated April 2, 1991.

iii. " Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," NUREG-0103, Revision 0, dated June 1, 1976. I

3. Surveillance Data Review:

A. The 18-month TS surveillance test results data for snubbers were reviewed for the period of the Fifth Refueling Outage (5RFO) through 9RFO. This time period was selected because it reflects the major plant improvements after June 1985, and covers five refueling outages and four cycles of test results.

B. The following ten snubbers were identified as failing the "as found" 10 percent functional test of Surveillance Requirement 4.7.7.2.c, for the time period. Note that eight of these snubbers failed in SRFO due to misadjusted bleed screws.

9RFO A 2.5 inch Accessible Grinnell snubber on hanger 33C-GCB-7-H4 on the Decay Heat line failed to lockup in l compression per the required acceptance range specified in I design drawing 12501-M-618. The snubber was disassembled to determine the cause of failure. A bent spring in the lockup compression barrel was the cause for the failed test. An engineering evaluation was performed, and it was determined the associated piping was within ASME Code allowable. An

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

l l e LAR 95-0019

( Enclosure 4 Page 3 i

additional 10% of the Grinnell Accessible Group was tested satisfactorily as required by Technical Specifications. The snubber was rebuilt with a new spring and was acceptable.

8RFO  !

No failures occurred in 10% functional sample.

l 7RFO l No failures occurred in 10% functional sample. l 6RFO A 3.25 inch Inaccessible Grinnell snubber located on hanger SR3 South on the main steam piping, failed to lockup in tension within the required acceptance range specified in design drawing 12501-M-618. The snubber locked up at an average of 5.3 IPM, but the acceptance criteria for a group l III Grinnell snubber is 6-23 IPM. This snubber fell I slightly short of the lockup criteria, most likely due to l piston rod wear which affected lockup velocity. An additional 10% Inaccessible Grinnell snubbers were tested, l with no additional failures, as required by Technical Specification. Engineering performed an evaluation of the j piping, and determined the loads and stresses were within ASME Code allowables.

5RFO l Six 1.5 inch Inaccessible Grinnell snubbers removed for functional testing, failed the as-found 10% test due to misadjustment of bleed and lockup adjusting screws. All l'.5 inch Inaccessible Grinnell snubbers were subsequently removed and functionally tested. Two more 1.5 inch snubbers ,

failed during the additional testing, for a total of eight j snubbers. All the snubbers removed were reset to the l appropriate lockup and bleed rater.

The following six snubbers were determined to be acceptable and did not impact the piping after engineering evaluation per TS 3/4.7.7:

l Acceptance As Found Failure Criteria 30-GCC-8-H1 East Bleedrate tension 11.5 maximum 10 30-GCC-8-H1 West Bleedrate tension 18.53 maximum 10 34-HCB-4-H34 Bleedrate compression 12.13 maximum 10 30-GCC-8-H16 Top Bleedrate tension 10.97 maximum 10 l

l 40-CCB-16-H6 Top Lockup Compression 5.86 minimum 6 l 1 40-CCD-16-H6 Bottom Lockup Compression 5.83 minimum 6 j I

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O LAR 95-0019 Enclosure 4 Page 4 I

Two 1.5 inch snubbers (34-HCD-4-H27 and 34-HCB-4-H34) were determined by engineering to be unacceptable after an engi-neering evaluation per TS 3/4.7.7 was performed. Engineer-  ;

ing determined the two snubbers would not have performed l their intended safety function. A safety evaluation deter- l

! mined the Containment Spray piping and pipe supports were  !

l within the USAR and ASME Code allowable limits. The failures were attributed to incorrectly adjusting the lockup and bleed rate adjusting screws, acceptance As Found criteria ]

l 34-HCB-4-H27 Lockup tension 24.2 maximum 23 i Lockup compression 24.1 maximum 23

Bleedrate tenrion 0 minimum 0.5 Bleedrate compression 0 minimum 0.5 '

l 34-HCD-4-H34 Lockup tension 26.3 maximum 23 Bleedrate tension 0 minimum 0.5 l C. Based on a review of the 18-month surveillance test results data,  !

no additional actions are necessary or recommended to support the  :

l increase in the present surveillance interval.  !

D. Based on the good overall historical performance and improved I performance of hydraulic snubbers from SRFO to the present, the ,

low potential for significant increases in failure rates of i hydraulic snubbers, and the introduction of no new failure modes, ,

it is concluded that it is acceptable to increase the surveillance  ;

interval for SR 4.7.7.2.b from 18 to 24-months, and that there is i as adverse effect on nuclear safety. Also it is acceptable to [

allow the continued application of TS 4.0.2 on a non-routine basis.

Should an inoperable snubber be identified under SR 4.7.7.2.b, SR l 4.7.7.2.d requires an escalation in the inspection sample until no further failures are found or until all such snubbers have been  ;

tested. l E.

References:

1. DBNPS Procedure ST-5044.01 " Inspection of Safety Related l Hydraulic Snubbers" i

11. DBNPS Procedere DB-MM-03006 " Inspection of Technical  !

l- Specification Hydraulic Snubbers" l .

l 111. DBNPS Drawing 12501-M-618 " Piston Settings, Locking l l Velocities, and Bleed Rates for Hydraulic Snubbers" {

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! 4. Maintenance Records Reviews  !

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l A. The maintenance records for TS snubbers were reviewed for the period of SRFO through 9RFO. This time period was selected i i f i

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O LAR 95-0019 l Enclosure 4 i Page 5 because it reflects the major plant improvements after June 1985, i

and covers five refueling outages and four operating cycles of maintenance activities.

I B. The maintenance records indicate the following failures are significant degradation cases during this time periods ,

l 9RFO I A failure on hanger 33C-GCB-7-H4 due to a bent spring, as discussed above.

8RFO Snubber removed from SR4 on Main Steam Line in the West D-ring was found to have heat discoloration and some fluid leakage. The snubber was tested satisfactorily and replaced l with another snubber.

7RFO The 2.5 inch Grinnell snubber removed from hanger 30-GCC-8-H9 on the pressurizer relief line was tested and the bleed rate in tension (5.31 ipm) was found to be slightly lower than the 12501-M-618 anubber specifications for a group IV (6.0 ipm minimum). The snubber had not been rebuilt since installation in the plant. Upon disassembly it was discovered that orifice plugs that were procedurally to be removed, were still installed, and a brass shaving had partially obstructed the orifice hole. An engineering l evaluation was performed, and it was determined the bleed l rate "as found" was very conservative in the tension bleed rate mode, and the snubber test "as found" was acceptable, i The snubber was rebuilt with the orifice plugs removed and reinstalled.

I l l The 1.5 inch Grinnell snubber removed from hanger PS-H31 on I the pressurizer spray line was found with the reservoir l cracked and empty during the visual inspection, and was part ,

l of the 10% functional test. The snubber was functionally i tested and passed. The Grinnell snubber was rebuilt as a spare, and a Lisgas hydraulic snubber which has an integral reservoir and is no+ as susceptible to heat, was installed in hanger PS-H31.

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The 1.5 inch Grinnell snubber removed from hanger PS-H34 on the pressurizer spray line was found with the reservoir I cracked and empty during the visual inspection, and was part of the 10% functional test. The snubber was functionally

tested and passed. The Grinnell snubber was rebuilt as a

{ spare, and a Lisega hydraulic snubber which has an integral

reservoir and is less susceptible to heat damage, was j installed in hanger PS-H34.

1 i The 1.5 inch Grinnell snubber removed from hanger PS-H3 on j the pressurizer spray line was found with the adjusting I

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s P 6 LAR 95-0019 Enclosure 4 Page 6 screws broken off the valve block of the snubber during visual inspection. The snubber was functionally tested and passed. The Grinnell snubber was rebuilt as a spare, and a Lisega snubber was installed in hanger PS-H3. I 1

The 1.5 inch Grinnell snubber removed from hanger 30-GCC-8-H4 on the pressurizer spray line was found with the reservoir cracked and empty during the visual inspection.

The snubber was functionally tested and passed. The Grinnell snubber was rebuilt and reinstalled.

The 2.5 inch Grinnell snubber removed from hanger  !

33B-CCB-6-H10 on Decay Heat train #2 line was found with  ;

inadequate fluid level during visual inspection. The I anubber was functionally tested and failed. It was determined upon investigation that an adjusting screw thread seal for bleedrate was leaking due to a loose thread seal

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locknut, which was not tightened properly when rebuilt in SRFO. A spare snubber was installed, and the removed snubber was rebuilt. An engineering evaluation / calculation j was performed and determined this condition would not have prevented the system from performing its intended function.

6RFO l l

The 1.5 inch Grinnell snubber removed from hanger 30-GCC-8-H1 on pressurizer relief line was found with an empty reservoir during a visual inspection. The snubber was functionally tested and passed. The adjusting screws leaked allowing the fluid to go low. The snubber was rebuilt, tested satisfactorily and reinstalled.

1 i

The two 6 inch Grinnell snubbers removed from hanger SR-8 in j the west D-ring on the main steam line were found with  !

empty reservoirs during a visual inspection. The snubbers I were functionally tested and passed. The snubbers were l l disassembled and found to have the fluid and seals in relatively good condition. The snubbers were replaced with ,

hydraulic Lisega snubbers. j j 5RFO 1

The upper 3.25 inch Grinnell snubber removed from hanger l EBB-1-SR4 on the main steam line was found with the

reservoir melted during a visual inspection. The snubber l was removed and functionally tested and failed. An engineering evaluation / calculation determined the piping was still within ASME Code allowables. The snubber was rebuilt with a metal reservoir to prevent recurrence.

Four Pacific Scientific (PSA) mechanical snubbers were determined to be inoperable while implementing a plant

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LAR 95-0019 Enclosure 4 Page 7 {

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change. Technical Specifications did not require testing of l mechanical'anubbers at that time. All mechanical snubbers  !

were replaced with Lisega hydraulic snubbers during SRFO. l C. Based on a review of the 18-month maintenance records, no addi-l l

tional actions are necessary or recommended to support this increase in the present surveillance interval.  !

D. As previously mentioned, the DBNPS has replaced the Pacific Scientific mechanical snubbers with Lisega hydraulic anubbers.

Additionally, in certain areas where conditions warrant, the installed Grinnell snubbers were replaced with Lisega hydraulic snubbers. The Lisega anubbers provide an improved service life and increased confidence of continued operability. The DBNPS has a good snubber operating history and the snubber degradation noted above is a small percentage of the total snubber population. Therefore, based on the good historical performance and improved performance of hydraulic snubbers from SRFO to the present, the low potential for significant increases in failure rates of hydraulic snubbers, and the introduction of no new l failure modes, it is concluded that it is acceptable to increase l the surveillance interval for SR 4.7.7.2.b from 18 to 24-months, and that there is no adverse effect on nuclear safety. Also it is acceptable to allow the continued application of TS 4.0.2 on a non-routine basis.

l E.

References:

The following maintenance records were used:

1. DBNPS Maintenance Work Order Records.

ii. DBNPS Preventive Maintenance Records.

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