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Amend 233 to License NPF-3,relocating TS Surveillance Requirement from TS 3/4.6.5.1, Shield Building - Emergency Ventilation Sys to TS Section 3/4.6.5.2, Shield Building Integrity
	ML20212H990
Person / Time
Site:	Davis Besse
Issue date:	06/22/1999
From:	Anthony Mendiola; NRC (Affiliation Not Assigned)
To:	;
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ML20212H994	List: ML20212H990; ML20212H996;
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	NUDOCS 9906280265
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p UNITE) STATES g

j NUCLEAR REGULATORY COMMISSION p

2 WASHINGTON, D.C. 20se6 c001 l

l FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO.1 l

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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No233 License No. NPF-3 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Toledo Edison Company, Centerior Service Company, and The Cleveland Electric illuminating Company (the licensees on the date of application; FirstEnergy Nuclear Operating Company became the sole licensed operator on January 1,1999) dated October 27,1998, as revised by letter of March 19,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as l

indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:

9906280265 990622 yDR ADOCK 05000346 PDR t

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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.233, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented not later than 120 days after issuance.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION l

'A.ony J. Mendiola, Chief, Section 2 roject Directorate ill Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date ofissuance: June 22, 1999 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 233 FACILITY OPERATING LICENSE NO. NPF-3 4

DOCKET NO. 50-346 i

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove triggi I

I 1-5 1-5 3/4 6-29 3/4 6-29 3/4 6-30 3/4 6-30 3/4 6-31 3/4 6-31 3/4 6-33 3/4 6-33 B 3/4 64 B 3/4 64 8 3/4 6-5 8 3/4 6-5

INDEX l

o DEFINITIONS SECTION PAGE JA_pEFINITIONS D EFIN E D TE R MS........................................

1-1 TH ERMAL POW ER.......................................

1-1 RATED THERMAL POWER.................................

1-1 OPERATION AL MODE....................................

1-1 A C TI O N................................................

1-1 OPER ABLE - OPERABluTY.................................

1-1 R EPORTAB LE EV ENT.....................................

1-2 CONTAINMENT INTEGRITY................................

1-2 CHANN EL CAUB RATION..................................

1-2 CHAN N EL CH EC K.......................................

1-2 CHANNEL FUNCTIONAL TEST..............................

1-3 CO R E ALTER ATIO N......................................

1-3 SHUTDOWN MAR GIN....................................

1-3 IDENTIFIED LEAKAGE....................................

1-3 U NIDENTIFIED LEAK AG E..................................

1-4

. PRESSURE BOUNDARY LEAKAGE...........................

1-4 CONTR OLLED LEAKAGE.................................

1-4 Q UADR ANT POWER TILT..................................

1-4 DOS E EQUIVALENT I-131..................................

1-4 E-AVERAGE DISINTEGRATION ENERGY.....................

1-4 STAGG ER ED TEST BASIS..................................

1-5 FR EQU ENCY NOTATION.................................

1-5 AXIAL POWER IMBALANCE...............................

1-5 D E L ET E D...............................................

l REACTOR PROTECTION SYSTEM RESPONSE TIME..............

1-5 SAFETY FEATURE RESPONSE TIME..........................

1-6

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PH YS ICS TESTS..........................................

1-6 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME 1-6 l

PROCESS CONTROL PROGRAM...........................

1-6a l

DAVIS-BESSE, UNIT 1 I

Amendment No. 38. 86,93,170, 233

i-DEFINITIONS j

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per disintegration (in MeV) for isotopes, other than lodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

STAGGERED TEST BASIS l'.21 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains or designated componen4 obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or designated components at the beginning of each subinterval.

FREQUENCY NOTATION E

1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

AXIAL POWER IMBALANCE -

1.23 AXlAL POWER IMBALANCE shall be the THERMAL POWER in the top half of the core expressed as a percentage of RATED THERMAL POWER minus the THERMAL POWER in the bottom half of the core expressed as a percentage of RATED THERMAL POWER.

1.24 DELETED l

REACTOR PROTECTION SYSTEM RESPONSE TIME

.1.25 The REACTOR PROTECTION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until power interruption at the control rod drive breakers.

DAVIS-BESSE, UNIT 1 1-5 Amendment No. 233

p CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l 1.

Verifying that the cleanup system satisfies the in-place penetration 9

L and bypass leakage testing acceptance criteria of less than 1%

l and uses the test procedure guidance in Regulatory Positions C.S.o, 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 days offer removal, that a laboratory analysis of a representative carbon sample obfoined in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory te: ting criteria of Regulatory Position C.6.o* of Regulatory Guide 1.52, Revision 2, March 1978, for o 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 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal odsorber operation by verifying, within 31 days offer removol, that a laborofory analysis of a representative torbon 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.o* of Regulatory Guide 1.52, Revision 2, March 1978, for a methyliodido penetration of less than 1%.

d.

At least once each REFUELING INTERVAL by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gouge while operating the system of a flow rate of 8,000 cfm i 10%:

2.

Verifying that the system starts automatically on any containment isolation test signal: and l

3.

Verifying that the filter cooling bypass volves con be monvolti opened.

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

DAVIS-BESSE, UNIT 1 3/4 6-29 Amendment No. 43,135,155,209,217,233

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CONTAINMENT SYSTEMS

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l SURVEILLANCE REQUIREMENTS (Continued) 4.

Deleted l

e.

After 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 occordance with ANSI N510-1980 for a DOP test oerosol while operating the system at a flow rate of 8000 cfm i 10%.

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After 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 hologenated hydrocarbon refrigeront test gas while operating the system at a flow rate of 8000 cfm i 10%.

DAVIS-BESSE, UNIT 1 3/4 6-30 Amendment No. 3,135,155,233

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CONTAINMENT SYSTEMS

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SHIELD BUILDING INTEGRITY l

l UMITING CONDITION FOR OPERATION 3.6.5.2 Shield building integrity shall be maintained.

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

Without shield building integrity, restore shield building integrity within 24 l

hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within'the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

'4.6.5.2.1 Shield building integrity shall be demonstrated at least once per l

31 days by verifying that airtight doors and the blowout panels listed in Table 4.6-1 are closed except'when the airtight doors are being used for normal transit entry and exit.

J' 4.6.5.2.2 Shield building integrity shall be demonstrated at least once per l

REFUELING INTERVAL by verifying that each Emergency Ventilation System l

train' produces a negative pressure of greater than or equal to 0.25 inches l

Water Gauge in the annulus within 4 seconds offer the fan attains a flow l.

rate of 8000 cfm 10%. This test is to be performed with the flow path l

established prior to starting the EVS fan, and the other dampers l

ossociated with the negative pressure boundary closed.

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I DAVIS-BESSE, UNIT 1 3/46-31 Amendment No.233

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TABLE 4.6-1 ACCESS OPENINGS REQUIRED TO BE CLOSED TO ENSURE SHIELD BUILDING INTEGRITY 1.

AIR TIGHT DOORS DOORNO.

DESCRIPTION ELEVATION 100 Access Door from the No.1 ECCS Pump Room 545' (Room 105) to Pipe Tunnel 101 104A Access Door from Stair AB-3 to the No.1 555' ECCS Pump Room (Room 105) 105 Access Door from Passage 110A to the creo above 555' the Decay Heat Coolers j

107 Access Door from the No. 2 ECCS Pump Room 555' (Room 115) to the Miscellaneous Waste Monitor Tonk and Pump Room (Room 114) 108 Access Door from the No. 2 ECCS Pump Room 555' (Room 115) to the Detergent Waste Drain Tank and Pump Room (Room 125) 201-A Access Door from Corridor 209 to the No.1 565' Mechanical Penetration Room (Room 208) 204 Access Door from Passage 227 to the Makeup 565' Pump Room (Room 225) 205 Access Door from Passage 227 to the No. 2 565' Mechanical Penetration Room (Room 236) 307 Access Door from Corridor 304 to the No. 3 585' Mechanical Penetration Room (Room 303) l 308 Access Door from Corridor 304 to the No. 4 585' Mechanical Penetration Room (Room 314) 11.

BLOWOUT PANELS TOTAL NO.

LOCATION ELEVATION 1

No. 2 Mechanical Penetration Room 565' (Room 236) 6 No. 3 Mechanical Penetration Room 585' (Room 303) 6 No. 4 Mechanical Penetration Room 585*

(Room 314)

DAVIS-BESSE, UNIT 1 3/4 6-33 Amendment No. 233

CONTAINMENT SYSTEMS BASES l

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3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the Hydrogen Analyzers, Containment Hydrogen i

Dilution System, and Hydrogen Purge System ensures that this equipment will be available to maintain the maximum hydrogen concentration within the containment vessel of or below inree volume percent following a LOCA.

The two redundant Hydrogen Analyzers determine the content of hydrogen within the coritainment vessel. The Hydrogen Analyzers, although they -

have their OPERABILITY requirements in this Specification, are considered port of the post-accident monitoring instrumentation of Specification 3/4.3.3.6, Post-Accident Monitoring instrumentation.

The Containment Hydrogen Dilution (CHD) System consists of two full capacity, redundant, rotary, positive' displacement type blowers to supply air to the containment. The CHD System controls the hydrogen concentration by the addition of air to the containment vessel, resulting in a pressurization of the containment'ond suppression of the hydrogen volume fraction.

~ Ee ConhbinmeniHidrogdPurge Syshis Filter Unit functions in T

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conjunction with the CHD System and is designed to release air from the containment atmosphere through a HEPA filter and charcool filter prior to discharge to the station vent.

As a backup to the CHD System and the Containment Hydrogen Purge I

System, the capability to install on external hydrogen recombination system has been provided.

3/4.6.5 SHIELD BUILDING 3/4.6.5.1 EMERGENCY VENTILATION SYSTEM The OPERABILITY of the emergency ventilation systems ensures that containment vessel leakoge occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal odsorber trains prior to discharge to the otmosphere. This requirement is necessory 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.

The proper functioning of the EVS fans, dampers, filters, adsorbers, etc., as l

o system is verified by the ability of each train to produce the required l

system flow rate.

l DAVIS-BESSE, UNIT 1 83/46-4 Amendment No. 66,167,183.233 Correction Lir dated 5/17/94 l

l CONTAINMENT SYSTEMS BASES 3/4.6.5.2 SHIELD BUILDING INTEGRITY Shield building integrity ensures that the release of radioactive material l

from the containment vessel _will be restricted to those leakage paths and associated leak rates assumed in the safety analysis. The closure of the airtight doors and blowout panels listed in Table 4.6-1 ensure that the Emergency Ventilation System (EVS) con provide a negative pressure between 0.25 and 1.5 inches Water Gauge within the annulus between the shield building and containment vessel and within the interconnecting mechanical penetration rooms after a loss-of-coolant accident (LOCA). This restriction,in conjunction with the operation of the EVS, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

In the event shield building integrity, including the capability of the EVS to l

provide a negative pressure of greater than or equal to 0.25 inches Water l

Gauge,is not maintained, shield building integrity must be restored within 24 l

hours. Twenty-four hours is a reasonable completion time considering the limited l

leakage design of the containment and the low probability of a Design Basis l

Accident occurring during.this time period.

l 3/4.6.5.3 SHIELD BUILDING STRUCTURAL INTEGRITY Deleted DAVIS-BESSE, UNIT 1 B 3/4 6-5 Amendment No. 135,205. 233 1

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