Text

Amends 183 & 64 to Licenses DPR-66 & NPF-73,respectively, Revise App a TSs Re Rc Leakage & Leakage Detection Sys in Effort to Bring TS Sections 3/4.4.6.1 & 3/4.4.6.2 Closer to NRC Improved Std TSs
	ML20073G538
Person / Time
Site:	Beaver Valley
Issue date:	09/22/1994
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20073G542	List: ML20073G538; ML20073G544;
References
	NUDOCS 9410040246
	Download: ML20073G538 (65)
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UNITED STATES l '^

S NUCLEAR REGULATORY COMMISSION 5

WASHINGTON. D.C. 205t&0001

/

DUOVESNE LIGHT COMPANY OHIO EDIS0N COMPANY PENNSYLVANIA POWER COMPANY DOCKET N0. 50-334 BEAVER VALLEY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.183 License No. DPR-66 1.

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

A.

The application for amendment by Duquesne Light Company, et al. (the licensee) dated June 2, 1994, 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.

9410040246 940922 PDR ADOCK 05000334 P

PDR

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License do. DPR-66 is hereby

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amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 183, are hereby incorporated in the license.

I The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

,ki

('

V Walter R. Butler, Director i

Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications l

Date of Issuance: September 22, 1994 l

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f

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ATTACHMENT TO LICENSE AMENDMENT NO.183 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert I

I VI VI VII VII XIII XIII 1-3 1-3 1-4 1-4 3/4 4-11 3/4 4-11 3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 5-8 B 3/4 4-2a B 3/4 4-2a B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a B 3/4 4-3b B 3/4 4-3c B 3/4 4-3d B 3/4 4-3e B 3/4 4-3f B 3/4 4-3g B 3/4 4-3h B 3/4 4-31 B 3/4 4-3j B 3/4 4-4 B 3/4 4-4 B 3/4 5-3 B 3/4 5-4 B 3/4 5-5

n DPR-66 INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS Defined Terms.

1-1 Thermal Power.

1-1 Rated Thermal Power.

1-1 Operational Mode 1-1 Action 1-1 operable - operability 1-1 Reportable Event 1-2 Containment Integrity.

1-2 Channel Calibration.

1-2 Channel Check.

1-2 Channel Functional Test.

1-3 Core Alteration.

1-3 Shutdown Margin.

1-3 Leakage 1-3 l

j Quadrant Power Tilt Ratio.

1-4 Dose Equivalent I-131 1-4 l

r Staggered Test Basis 1-4

)

Frequency Notation 1-4 I

Reactor Trip System Response Time.

1-5 1

Engineered Safety Feature Response Time.

1-5 i

BEAVER VALLEY - UNIT 1 I

Amendment No.183 r

DPR-66 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE j

i 3/4.4 REACTOR COOLANT SYSTEM i

3/4.4.1 REACTOR COOLANT LOOPS 3/4.4.1.1 Normal Operation 3/4 4-1 3/4.4.1.2 Hot Standby.

3/4 4-2b 3/4.4.1.3 Shutdown 3/4 4-2c 3/4.4.1.4 Isolated Loop.

3/4 4-3 3/4.4.1.5 Isolated Loop Startup.

3/4 4-4 3/4.4.1.6 Reactor Coolant Pump Startup 3/4 4-4a 3/4.4.2 SAFETY VALVES - SHUTDOWN 3/4 4-5 3/4.4.3 SAFETY VALVES - OPERATING 3/4 4-6 3/4.4.4 PRESSURIZER.

3/4 4-7 3/4.4.5 STEAM GENERATORS 3/4 4-8 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakage Detection Instrumentation.

3/4 4-11 l

3/4.4.6.2 Operational Leakage 3/4 4-13 3/4.4.6.3 Pressure Isolation Valves 3/4 4-14a l

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3/4.4.7 CHEMISTRY 3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY.

3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 3/4.4.9.1 Reactor Coolant System 3/4 4-22 BEAVER VALLEY - UNIT 1 VI Amendment No.183

DPR-66 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9.2 Pressurizer.

3/4 4-27 3/4.4.9.3 Overpressure Protection Systems 3/4 4-27a 3/4.4.10 STRUCTURAL INTEGRITY - ASME Code Class 1,

2 and 3 Components.

3/4 4-28 3/4.4.11 RELIEF VALVES 3/4 4-29 3/4.4.12 REACTOR COOLANT SYSTEM VENTS 3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,, 2 350'F.

3/4 5-3 m

3/4.5.3 ECCS SUBSYSTEMS - T,, < 350*F.

3/4 5-6 3/4.5.4 BORON INJECTION SYSTEM 3/4.5.4.1.1 Boron Injection Tank 2 350*F

. 3/4 5-7 3/4.5.4.1.2 Boron Injection Tank < 350*F 3/4 5-7a 3/4.5.5 SEAL INJECTION FLOW 3/4 5-8 l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 Containment Integrity.

3/4 6-1 3/4.6.1.2 Containment Leakage.

3/4 6-2 3/4.6.1.3 Containment Air Locks.

3/4 6-5 3/4.6.1.4 Internal Pressure.

3/4 6-6 3/4.6.1.5 Air Temperature.

3/4 6-8 3/4.6.1.6 Containment Structural Integrity 3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 Containment Quench Spray System.

3/4 6-11 BEAVER VALLEY - UNIT 1 VII Amendment No.183 i

DPR-66 INDEX BASES SECTION PAGE 3/4.3.3.7 Chlorine Detection Systems B 3/4 3-3 l

3/4.3.3.8 Accident Monitoring Instrumentation.

B 3/4 3-3 3/4.3.3.9 Radioactive Liquid Effluent Monitoring Instrumentation.

B 3/4 3-4 3/4.3.3.10 Radioactive Gaseous Effluent Monitoring f

Instrumentation.

B 3/4 3-4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS B 3/4 4-1 f

3/4.4.2 and 3/4.4.3 SAFETY VALVES B 3/4 4-la 3/4.4.4 PRESSURIZER.

B 3/4 4-2 3/4.4.5 STEAM GENERATORS B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-3 i

3/4.4.6.1 Leakage Detection Instrumentation.

B 3/4 4 l 3/4.4.6.2 Operational Leakage.

B 3/4 4-3e l

3/4.4.7 CHEMISTRY.

B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.

B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY B 3/4 4-10 3/4.4.11 RELIEF VALVES B 3/4 4-10 3/4.4.12 REACTOR COOLANT SYSTEM VENTS B 3/4 4-11 i

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3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS B 3/4 5-1 I

B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS

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3/4.5.4 BORON INJECTION SYSTEM B 3/4 5-1 3/4.5.5 SEAL INJECTION FLOW B 3/4 5-3 l

BEAVER VALLEY - UNIT 1 XIII Amendment No.183

m DPR-66 i

DEFINITIONS I

CHANNEL FUNCTIONAL TEST E

i 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable i

to verify OPERABILITY including alarm and/or trip functions.

l CORE ALTERATION

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1.12 CORE ALTERATION shall be the movement or msnipulation of any

{

component within the reactor-pressure vessel with the vessel-head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall' not preclude completion of movement of a component to a safe conservative position.

i SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity' by which the reactor is or would be subcritical from its present l

condition assuming all full length rod cluster assemblies (shutdown-and control) are fully inserted except for the single rod cluster i

assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage-datection systems or not to be Pressure Boundary LEAKAGE, or 3.

Reactor coolant system LEAKAGE through a

steam generator to the secondary system.

b.

Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

BEAVER VALLEY - UNIT 1 1-3 Amendment No.183

)

O DPR-66 DEFINITIONS c.

Pr?ssure Boundary LEAKAGE Pressure Boundary LEAKAGE shall be LEAKAGE (except steam generator tube LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

1.15 THROUGH 1.17 (DELETED)

OUADAANT POWER TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one (1) excore detector inoperable, the remaining three (3) detectors shall be used for computing the average.

DOSE EOUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (yci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a.

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

The testing of one (1) system, subsystem, train or other designated component at the beginning of each subinterval.

FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

BE. DER VALLEY - UNIT 1 1-4 Amendment No.183

r DPR-66 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection instrumentation shall be OPERABLE:

a.

One containment sump (narrow range level or discharge flow) monitor; and b.

One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY:

MODES 1, 2,

3 and 4.

ACTION:

a.

With the required containment sump monitor inoperablem, operations may continue for up to 30 days provided that a Reactor Coolant System water inventory balance measurement

[

(Specification 4.4.6.2.b) is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, 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 .

With the reg)uired containment atmosphere radioactivity monitor b.

inoperable (

operations may continue for up to 30 days provided:

1.

Grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or 2.

A Reactor Coolant System water inventory balance measurement (Specification 4.4.6.2.b) is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, 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 .

(1) The provisions of Specification 3.0.4 are not applicable.

BEAVER VALLEY - UNIT 1 3/4 4-11 Amendment No.183

1 i

DPR-66 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

With the required containment sump monitor and the containment c.

atmosphere radioactivity monitor inoperable, 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 .

1 l

SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection instrumentation shall be demonstrated OPERABLE by:

j a.

Performance of a CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitor at the frequencies specified in Table 4.3-3.

b.

Performance of a

CHANNEL CALIBRATION of the required containment sump monitor at least once per 18 months.

BEAVER VALLEY - UNIT 1 3/4 4-12 Amendment No.183 e

DPR-66 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE, b.

1 gpm unidentified LEAKAGE, c.

1 gpm total primary to secondary LEAKAGE through all steam generators, d.

500 gallons per day primary to secondary LEAKAGE through any one steam generator, and e.

10 gpm identified LEAKAGE.

APPLICABILITY:

MODES 1, 2,

3, and 4.

ACTION:

a.

With any pressure boundary LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System LEAKAGE greater than any one of the above limits, excluding pressure boundary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 4.4.6.2 Reactor Coolant System LEAKAGES shall be demonstrated to be within each of the above limits by:

l a.

Monitoring the following leakage detection instrumentation at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:m 1.

Containment atmosphere gaseous radioactivity monitor.

(1) Only on leakage detection instrumentation required by LCO 3.4.6.1.

BEAVER VALLEY - UNIT 1 3/4 4-13 Amendment No.183

DPR-66 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 2.

Containment atmosphere particulate radioactivity monitor.

3.

Containment sump discharge flow monitor.

4.

Containment sump narrow range level monitor, b.

Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation.c2) l (2) Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

l BEAVER VALLEY - UNIT 1 3/4 4-14 Amendment No.183

o NPF-73 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be less than or equal to 28 gpm with the charging pump discharge pressure greater than or equal to 2410 psig and the seal injection flow control valve full open.

APPLICABILITY:

MODES 1, 2,

and 3.

ACTION:

With the seal injection flow not within the limit, adjust a.

manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure greater than or equal to 2410 psig and the seal injection flow j

control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT i

STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4 Verify at least once per 31 days that the valves are adjusted to give a flow within the limit with the charging pump discharge at greater than or equal to 2410 psig and the seal injection flow control j

valve full open.'"

(1) Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at greater than or equal to 2215 psig and less than or equal to 2255 psig.

1 BEAVER VALLEY - UNIT 2 3/4 5-7 Amendment No.64

DPR-66 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.5 Reactor coolant pump seal injection flow shall be less than or equal to 28 gpm with the charging pump discharge pressure greater than or equal to 2311 psig and the seal injection flow control valve full open.

l APPLICABILITY:

MODES 1, 2,

and 3.

ACTION:

a.

With the seal injection flow not within the limit, adjust manual seal injection throttle valves to give a flow within

{

the limit with the charging pump discharge pressure greater than or equal to 2311 psig and the seal injection flow control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.5 Verify at least once pur 31 days that the valves are adjusted to give a flow within the limit with the charging pump discharge at greater than or equal to 2311 psig and the seal injection flow control valve full open.")

l (1) Not required to be perforned until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at greater than or equal to 2210 psig and less than or equal to 2250 psig.

BEAVER VALLEY - UNIT 1 3/4 5-8 Amendment No.183

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DPR-66 REACTOR COOLANT SYSTEM i

BASES

]

J J

3/4.4.5 STEAM GENERATORS (Continued)

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operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary to secondary LEAKAGE = 500 gallons per day per steam generator).

Cracks having a primary to secondary LEAKAGE less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated l

accidents.

Operating plants have demonstrated that primary to i

secondary LEAKAGE of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. l Leakage in excess of this limit will require plant shutdown and an i

unscheduled inspection, during which the leaking tubes will be located and plugged.

l Wastage-type defe, cts are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type j

should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging or repair will be required of all tubes with imperfections exceeding the plugging or i

j repair limit.

Degraded steam generator tubes may be repaired by the j

installation of sleeves which span the degraded tube section. A steam generator tube with a

sleeve installed meets the structural i

requirements of tubes which are not degraded, therefore, the sleeve is l

considered a part of the tube. The surveillance requirements identify l

those slesving methodologies approved for use.

If an installed sleeve j

is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged.

The plugging limit for the i

sleeve is derived from R.G. 1.121 analysis which utilizes a 20 percent i

allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.

Steam generator tube inspections of operating plants have demonstrated the i

capability to reliably detect degradation that has penetrated 20 i

percent of the original tube wall thickness.

l l

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to i

the Commission pursuant to Specification 6.6 prior to resumption of j

plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, j

laboratory examinations, tests, additional addy-current inspection, and revision of the Technical Specifications, if necessary.

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l BEAVER VALLEY - UNIT 1 B 3/4 4-2a Amendment No.183 g,m..

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DPR-66 REACTOR COOLANT SYSTEM I

BASES 3 /4. 4. 6.1 TRAKAGE DETECTION INSTRUMENTATION l

BACKGROUND GDC 30 of Appendix A to 10 CFR 50 requires means for detecting and, to f

the extent practical, identifying the source of RCS LEAKAGE.

Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure.

Thus, an early indication or warning I

signal is necessary to permit proper evaluation of all unidentified LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm I

can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump.

The non-ECCS portion of the containment sump used to collect unidentified LEAKAGE is instrumented to alarm due to abnormal increases in the water inventory.

The sensitivity is acceptable for detecting increases in unidentified LEAKAGE.

The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial l

reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE.

An increase in humidity of the containment atmosphere would indicate i

release of water vapor to the containment.

Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE.

Since the humidity level is influenced by several

factors, a

quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump.

Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem.

Humidity monitors are not required by this LCO.

1 BEAVER VALLEY - UNIT 1 B 3/4 4-3 Amendment Noj83

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DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

BACKGROUND (Continued) l Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment.

Containment temperature and pressure fluctuate slightly during plant operation, j

but a rise above the normally indicated range of values may indicate RCS leakage into the containment.

The relevance of temperature and pressure measdrements are affected by containment free volume and, for temperature, detector location.

Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and pressure monitors are not required by this LCO.

j APPLICABLE SAFETY ANALYSES The need to evaluate the severity of an alarm or an indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary.

Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration.

Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakage occur detrimental to the safety of the unit and the public.

LC.Q One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks.

This i

LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.

l BEAVER VALLEY - UNIT 1 B 3/4 4-3a Amendment No.183

i DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

LCO (Continued)

The LCO is satisfied when monitors of diverse measurement means are available.

Thus, the containment sump monitor, in combination with a i

gaseous or particulate radioactivity monitor, provides an acceptable minimum. The containment sump monitor is comprised of the instruments i

associated with the non-ECCS portion of the containment sump which j

monitor narrow range level and sump pump discharge flow.

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2, 3, and 4,

RCS leakage detection instrumentation is required to be OPERABLE.

i In MODE 5 or 6, the temperature is to be less than or equal to 200*F and pressure is maintained low or at atmospheric pressure.

Since the temperatures and pressures are far lower than those for MODES 1, 2, 3, l

and 4,

the likelihood of leakage and crack propagation are much j

smaller.

Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

i ACTIONS i

A.

With the required containment sump monitor inoperable, no other form of sampling can provide the equivalent information; i

however, the containment atmosphere radioactivity monitoring system will provide indications of changes in leakage.

Together with the atmosphere

monitor, the periodic surveillance for RCS water inventory balance, SR 4.4.6.2.b, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage.

Restoration of the required sump monitor to OPERABLE status within a completion Time of 30 days is required to regain the function after the monitor's failure.

This time is acceptable, considering the frequency and adequacy of the RCS water inventory balance required by Required Action "a."

BEAVER VALLEY - UNIT 1 B 3/4 4-3b Amendment No.183 i

DPR-66 REACTOR COOLANT SYSTEM BASES

.2/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

ACTIONS (Continued)

Required Action "a" is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is allowed when the containment sump monitor is inoperable.

This allowance is provided becauss other instrumentation is available to monitor RCS leakage, b.1.

and b.2.

With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with SR. 4.4.6.2.b, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of the required containment atmosphere radioactivity monitors.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage.

The 30 day Completion Time recognizes at least one other form of leakage detection is available, l

Required Action "b" is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is allowed when the gaseous and particulate containment atmosphere radioactivity monitor channel is inoperable.

This allowance is provided because other instrumentation is available to monitor for RCS LEAKAGE.

s.

With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown is required.

The plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems.

BEAVER VALLEY - UNIT 1 B 3/4 4-3c Amendment No.183 i

\\

1 DPR-66.

REACTOR COOLANT SYSTEM i

i BASES i

3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION fContinuedi SURVEIT.TANCE REOUIREMENTS (SR) j SR 4.4.6.1.a I

i i

SR 4.4.6,1.a requires the performance of a CHANNEL CHECK of the l

required containment atmosphere radioactivity monitor.

The check gives reasonable confidence that the channel is operating properly.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

I SR 4.4.6.1.a requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitor.

The test ensures that the monitor can perform its function in the desired The test. verifies the alarm setpoint and relative accuracy of manner.

the instrument string.

The Frequency of 31 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

SR 4.4.6.1.a also requires the performance of a CHANNEL CALIBRATION on the required containment atmosphere radioactivity monitor.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The Frequency of 18 months is a typical refueling cycle and considers channel reliability.

Again, operating experience has proven that this Frequency is l

acceptable.

1 i

I SR 4.4.6.1.b

)

.i SR 4.4.6.1.b requires the performance of a CHANNEL CALIBRATION on the j

required containment sump monitor.

The calibration verifies the j

accuracy of the instrument string, including the instruments located j

inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability.

Again, operating experience i

has proven that this Frequency is acceptable.-

3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND j

components that contain or transport the coolant to or from the reactor core make up the RCS.

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

BEAVER VALLEY - UNIT 1 B 3/4 4-3d Amendment No.183 l

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

BACKGROUND (Continued)

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety.

This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE.

Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration.

Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100 percent leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE.

However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can BEAVER VALLEY - UNIT 1 B 3/4 4-3e Amendment NoJ83

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued) affect the probability of such an event.

The safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes a 10 gpm primary to secondary LEAKAGE.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR).

The leakage contaminates the secondary fluid.

The SLB is more limiting for site radiation releases.

The safety analysis for the SLB accident conservatively assumes a 10 gpm primary to secondary LEAKAGE.

The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits).

LC2 RCS operational LEAKAGE shall be limited to:

a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the RCPB.

LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Should pressure boundary LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

BEAVER VALLEY - UNIT 1 B 3/4 4-3f Amendment No.183

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued),

LCO (Continued) b.

Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c.

Primary to Secondary LEAKAGE throuch All Steam Generators (SGs)

Total primary to secondary LEAKAGE amounting to 1 gpm through all SGs produces acceptable offsite doses in the SLB accident analysis. Violation of this LCO could exceed the offsite dose limits for this accident.

Primary to secondary LEAKAGE must be included in the total allowable limit for identified LEAKAGE.

d.

Primary to Secondary LEAKAGE throuch Any One SG The 500 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked through many cracks, the cracks are very

small, and the above assumption is conservative.

e.

Identified LEAKAGE Up to 10 gpm of" identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of identified LEAKAGE and is well within the capability of the RCS Makeup System.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amendment No.183

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABILITY In MODES 1, 2,

3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.6.3, "RCS Pressure Isolation Valve (PIV)," measures leakage through each individual PIV and can impact this LCO.

Of the two PIVs in series in each isolated line, leakage measured th: vagh one PIV does not result in RCS LEAKAGE when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS g.

If any pressure boundary LEAKAGE exists, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

The reactor must be brought to MODE 3

)

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This action 1

reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely, b.

Unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down.

This action is necessary to prevent further deterioration of the RCPB.

If the unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE BEAVER VALLEY - UNIT 1 B 3/4 4-3h Amendment No.183

o 1

DPR-66 P

REACTOR COOLANT SYSTEM BASES I

i 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS (Continued) cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

The t

reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This action reduces the LEAKAGE.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses l

acting on the RCPB are much lower, and further deterioration i

is much less likely.

SURVEILLANCE REOUIREMENTS (SR)

SR 4.4.6.2 Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.

It should be noted that LEAKAGE past seals i

and gaskets is not pressure boundary LEAKAGE.

Unidentified.. LEAKAGE l

and identified LEAKAGE are determined by performance of an RCS water inventory balance.

Primary to secondary LEAKAGE is also measured by l

performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure.

Therefore, j

this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure.have been j

established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note j

requires the Surveillance to be met when steady state is established.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

)

BEAVER VALLEY - UNIT 1 B 3/4 4-31 Amendment No.183

]

)

DPR-66 l

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REOUIREMENTS (SR) (Continued)

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring of the leakage detection system is sufficient to provide an early warning of increased RCS LEAKAGE. These leakage detection systems are specified in LCO 3.4.6.1,

" Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

Note (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance is required only on leakage detection instrumentation required by LCO i

3.4.6.1.

This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended on leakage detection instrumentation which is inoperable or not required to be operable per LCO 3.4.6.1.

Note (2) states that this SR is required to be performed during steady state operation.

3/4.4.6.3 PRESSURE ISOLATION VALVE' LEAKAGE The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure.

It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.

Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA'.

Leakage from the RCS pressure isolation valve is identified LEAKAGE l.

and will be considered as a portion of the allowed limit.

BEAVER VALLEY - UNIT 1 B 3/4 4-3j Amendment No.183 l

DPR-66 REACTIVITY CONTROL SYSTEMS BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant system over the life of the plant.

The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

BEAVER VALLEY - UNIT 1 B 3/4 4-4 Amendment No.183

DPR-66 r

l EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 SEAL INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the Emergency Cora Cooling Systems (ECCS) throttle valves in that each restricts flow from the charging pump header to the Reactor Coolant Systems (RCS).

The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident.

This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during SI.

APPLICABLE SAFETY ANALYSES All ECCS subsystems are taken credit for in the large break loss of coo'lant accident (LOCA) at full power.

The LOCA analysis establishes i

the minimum flow for the ECCS pumps.

The charging pumps are also credited in the small beak LOCA analysi.

This analysis establishes the flow and discharge head at the design point for the charging pumps.

The steam generator tube rupture and main steam line break event analyses also credit the charging pumps, but are not limiting in their design.

Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses.

This LCO ensures that seal injection flow of less than or equal to 28 gpm, with charging pump discharge pressure greater than or equal to 2311 psig and seal injection flow control valve full open, will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA.

It also ensures that the charging pumps will deliver sufficient water for a

small LOCA and sufficient boron to maintain the core subcritical.

For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory.

BEAVER VALLEY - UNIT 1 B 3/4 5-3 Amendment No.183

DPR-66 EMERGENCY CORE COOLING SYSTEMS I

3/4.5.5 SEAL INJECTION FLOW (Continued) m

)

The intent of the LCO limit on seal injection flow is to make sure I

that flow through the RCP seal water injection line is low enough to ensure that charging pump injection flow is directed to the RCS via i

the injection points in accordance with the requirements of 10 CFR 50.46.

The LCO itt net strictly a flow limit, but rather a flow limit based on i

a flow line resistance.

In order to establish the proper flow line resistance, a pressure and flow must be known.

The flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure and that the charging pump discharge pressure is greater than or equal to the value specified in this LCO.

The charging pump discharge pressure remains essentially constant through all the applicable MODES of this LCO..

A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure.

The valve settings established at the prescribed charging pump discharge pressure result in a conservative valve position should RCS pressure decrease.

The l

additional modifier of this LCO, the air operated seal injection control valve being full open, is required since the valve is designed to fail open for the accident condition.

With the discharge pressure and control valve position as specified by the LCO, a flow limit is established.

It is this flow limit that is used in the accident analyses.

1 The limit on seal injection flow, combined with the charging pump discharge pressure limit and an open wide condition of the seal injection flow control valve, must be met to render the ECCS OPERABLE.

i If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses.

l APPLICABILITY In MODES 1,

2, and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4.

The seal injection flow limit is not applicable for MODE 4 and lower because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES.

Therefore, RCP seal injection flow must be limited in MODES 1, 2,

and 3 to ensure adequate ECCS performance.

BEAVER VALLEY - UNIT 1 B 3/4 5-4 Amendment No.183

l DPR-66 D S GENCY CORE COOLING SYSTEMS BASES 3/4.5.5 SEAL INJECTION FLOW (Continued)

ACTIONS n.

With seal injection flow exceeding its limit, the amount of charging flow available to the RCS may be reduced.

Under this Condition, action must be taken to restore the flow to below its limit.

The operator has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the flow is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis.

The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and ensures that seal injection flow is restored to or below its limit.

This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel, When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.

Continuing the plant shutdown begun in this nequired Action, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE REOUIREMENTS (SR)

SR 3.5.5.1 Verification every 31 days that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained.

The Frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve surveillance Frequencies.

The Frequency has proven to be acceptable through operating experience.

As noted, the Surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized within a i 20 psig range of normal operating pressure.

The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.

The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the Surveillance is timely.

BEAVER VALLEY - UNIT 1 B 3/4 5-5 Amendment No.183 1

putsu f.-

UNITED STATES

~

NUCLEAR REGULATORY COMMISSION f

WASHit;GTON, D.C. 20SSH001

+.,

/

DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLED0 EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENS1 Amendment No. 64 License No. NPF-73 1.

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

A.

The application for amendment by Duquesne Light Company, et al. (the licensee) dated June 2, 1994, 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.

. ~_.

i

, i 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 64, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are i

hereby incorporated in the license. DLC0 shall operate the facility in accordance with the Technical Specifications and the l

Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION kNLl&

u v

Walter R. Butler, Director Project Directorate I-3 i

Division of Reactor Projects - I/II i

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: September 22, 1994

ATTACHMENT TO LICENSE AMENDMENT NO. 64 FACILITY OPERATING LICENSE NO. NPF-73 i

DOCKET NO. 50-412 i

Replace the following pages of Appendix A, Technical Specifications, I

with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert I

I II II VI VI XI XI 1-3 1-3 j

1-4 1-4 1-5 1-5 i

1-6 1-6 1-7 1-7 1-8 1-8 1-9 f

I 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 3/4 4-20 3/4 4-20 3/4 5-7 B 3/4 4-3 B 3/4 4-3 B 3/4 4-4 B 3/4 4-4 B 3/4 4-4a 5

B 3/4 4-4b B 3/4 4-4c B 3/4 4-4d B 3/4 4-4e B 3/4 4-4f B 3/4 4-4g B 3/4 4-4h B 3/4 4-41 B 3/4 4-4j B 3/4 4-5 B 3/4 4-5 B 3/4 5-2 B 3/4 5-3 B 3/4 5-4 1

+

h L

l

~

... ~ -

l i

4 NPF-73 INDEX DEFINITIONS SECTION PAGE i

1.0 DEFINITIONS 1.1 DEFINED TERMS 1-1 4

i 1.2 THERMAL POWER.

1-1 1.3 RATED THERMAL POWER.

1-1 l

l 1.4 OPERATIONAL MODE 1-1 r

i 1.5 ACTION 1-1 1.6 OPERABLE - OPERABILITY 1-1 l

1.7 REPORTABLE EVENT 1-1 1.8 CONTAINMENT INTEGRITY.

1-1 1.9 CHANNEL CALIBRATION.

1-2 1.10 CHANNEL CHECK 1-2

)

1.11 CHANNEL FUNCTIONAL TEST 1-2 1.12 CORE ALTERATION 1-2 i

1.13 SHUTDOWN MARGIN 1-3 1

1.14 LEAKAGE.

1-3 1.15 DELETED 1.16 DELETED 1.17 DELETED 1.18 QUADRANT POWER TILT RATIO 1-3 1.19 DOSE EQUIVALENT I-131 1-4 1.20 STAGGERED TEST BASIS 1-4 1.21 FREQUENCY NOTATION.

1-4 1.22 REACTOR TRIP RESPONSE TIME.

1-4 1.23 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-4 BEAVER VALLEY - UNIT 2 I

Amendment. No.64 F-73 i

i

t I

INDEX DEFINITIONS SECTION PAGE 1.24 AXIAL FLUX DIFFERENCE 1-4 1.25 PHYSICS TEST.

1-5 1.26 E-AVERAGE DISINTEGRATION ENERGY 1-5 1.27 SOURCE CHECK.

1-5 1.28 PROCESS CONTROL PROGRAM 1-5 1.29 SOLIDIFICATION 1-5 1.30 OFF-SITE DOSE CALCULATION MANUAL (ODCM) 1-5 1.31 GASEOUS RADWASTE TREATMENT SYSTEM 1-6 1.32 VENTILATION EXHAUST TREATMENT SYSTEM 1-6 1.33 PURGE - PURGING 1-6 1.34 VENTING 1-6 1.35 MAJOR CHANGES 1-6 1.36 MEMBER (S) OF THE PUBLIC 1-7 i

TABLE 1.1 OPERATIONAL MODES (TABLE 1.1) 2-8 i

TABLE 1.2 FREQUENCY NOTATION 1-P i

BEAVER VALLEY - UNIT 2 II Amendment No.64

NPF-73 INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Instrumentation.

3/4 4-17 l

Operational Leakage.

3/4 4-19 Pressure Isolation Valves.

3/4 4-21 3/4<4.7 CHEMISTRY.

3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY.

3/4 4-27 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System 3/4 4-30 Pressurizer.

3/4 4-34 Overpressure Protection Systems.

3/4 4-35 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components 3/4 4-38 3/4.4.11 RELIEF VALVES 3/4 4-39 3/4.4.12 REACTOR COOLANT SYSTEM HEAD VENTS 3/4 4-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T, 1 350*F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T, < 3 50

F 3/4 5-6 3/4.5.4 SEAL INJECTION FLOW.

3/4 5-7 l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.

3/4 6-1 Containment Leakage.

3/4 6-2 Containment Air Locks 3/4 6-4 Internal Pressure.

3/4 6-6 Air Temperature.

3/4 6-8 Containment Structural Integrity 3/4 6-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System.

3/4 6-10 BEAVER VALLEY - UNIT 2 VI Amendment No.64

O NPF-73 I DEX BASES SECTION PAGE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-4 3/4.4.7 CHEMISTRY

. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY.

B 3/4 4-15 7

3/4.4.11 RELIEF VALVES B 3/4 4-16 3/4.4.12 REACTOR COOLANT SYSTEM HEAD VENTS B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS B 3/4 5-1 3/4.5.2 AND 3/4.5.3 ECCS SUBSYSTEMS B 3/4 5-1 3/4.5.4 SEAL INJECTION FLOW B 3/4 5-2 l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES B 3/4 6-2 3/4.6.4 COMBUSTIBLE GAS CONTROL B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM.

B 3/4 6-3 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.

B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM.

B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM.

B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK B 3/4 7-3 3/4.7.6 FLOOD PROTECTION.

B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEM.

B 3/4 7-4 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)

B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION B 3/4 7-5 3/4.7.12 SNUBBERS B 3/4 7-5 BEAVER VALLEY - UNIT 2 XI Amendment No. 64

i e

NPF-73 1

DEFINITIONS LEAKAGE 1.14 LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection i

systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems j

or not to be Pressure Boundary LEAKAGE, or 3.

Reactor coolant system LEAKAGE through a steam generator to the secondary system.

b.

Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

c.

Pressure Boundarv LEAKAGE Pressure Boundary LEAKAGE shall be LEAKAGE (except steam generator tube LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

1.15 THROUGH 1.17 (DELETED)

OUADRANT POWER TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one (1) excore detector inoperable, the remaining three (3) detectors shall be used for computing the average.

BEAVER VALLEY - UNIT 2 1-3 Amendment No. 64

NPF-73 DEFINITIONS DOSE EOUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (yCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977 or TID 14844.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a.

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

The testing of one (1) system, subsyctem, train or other designated component at the beginning of each subinterval.

j i

i FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

REACTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required

values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector.

BEAVER VALLEY - UNIT 2 1-4 Amendment No.64

NPF-73 DEFINITIONS PHYSICS TESTS t

1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

5 - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

PROCESS CONTROL PROGRAM 1.28 A PROCESS CONTROL PROGRAM (PCP) shall be the manual or set of operating parameters detailing the program of sampling, analysis, and evaluation by which SOLIDIFICATION of wet radioactive wastes is assured. Requirements of the PCP are provided in Specification 6.14.

SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

OFFSITE DOSE CALCULATION MANUAL IODCM) 1.30 An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. Requirements of the ODCM are provided in Specification 6.15.

P.EAVER VALLEY - UNIT 2 1-5 Amendment No. 64

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NPF-73 DEFINITIONS l

GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

VENTILATION EXHAUST TREATMENT SYSTEM 1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose l

of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is. not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air

~

or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is required to purify the confinement.

VENTING 1.34 VENTING is the controlled process of discharging air or gas from a

confinement to maintain temperature,

pressure, humidity,.

concentration or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

MAJOR CHANGES 1.35 MAJOR CHANGES to radioactive waste systems, as addressed in Paragraph 6.16.2, (liquid, gaseous and solid) shall include the following:

1)

Major changes in process equipment, components, structures, and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the BEAVER VALLEY - UNIT 2 1-6 Amendment No.64

\\

- -. - - ~ - -

[

p NPF-73 DEFINITIONS A

MAJOR CHANGES (Continued) i staff's Safety Evaluation Report (SER)

(e.g., deletion of avaporators and installation of domineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems),

2)

Major changes in the design of radwaste treatment systems (liquid,

gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);

3)

Changes in system design which may invalidate the accident analysis as described in the SER (e.g.,

changes in tank capacity that would alter the curies released); and 4)

Changes in system design that could potentially result in a

significant increase in occupational exposure of i

operating personnel (e.g.,

use of temporary. equipment without adequate shielding provisions).

MEMBER (S) OF THE PUBLIC 1.36 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or its vendors.

Also excluded from this category are persons who enter. the site to service equipment or to make deliveries and persons who traverse portions of the site as the consequence of a public highway, railway, or waterway located within the confines of the site boundary. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

CORE OPERATING LIMITS REPORT 1.37 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.14. Plant operation within these operating limits is addressed in individual specifications.

l I

1 BEAVER VALLEY - UNIT 2 1-7 Amendment No. 64 l

l NPF-73

[

TABLE 1.1 OPERATIONAL MODES f

i REACTIVITY

% RATED AVERAGE CONDITION, THERMAL COOLANT MODE K.rc POWER

TEMPERATURE 1.

POWER OPERATION 20.99

>5%

2350*F L

2.

STARTUP 20.99

$5%

2350*F 3.

HOT STANDBY

<0.99 0

2350*F

4. HOT $HUTDOWN

<0.99 0

350*F >T,y,

>200*F 5.

COLD SHUTDOWN

<0.99 0

$200*F

6. REFUELING **

$0.95 0

$140*F

Excluding decay heat.

- Reactor vessel head unbolted or removed and fuel in the vessel.

BEAVER VALLEY - UNIT 2 1-8 Amendment No.64 l

NPF-73 TABLE 1.2 i

FREQUENCY NOTATION NOTATION FREQUENCY T

S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l D

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A.

Not applicable.

i i

BEAVER VALLEY - UNIT 2 1-9 Amendment No. 64 l

NPF-73 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection instrumentation shall be OPERABLE:

L a.

One containment sump (narrow range level or discharge flow) monitor; and b.

One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY:

MODES 1, 2,

3 and 4.

ACTION:

a.

With the required containment sump monitor inoperable (0, operations may continue for up to 30 days provided that a Reactor Coolant System water inventory balance measurement (Specification 4.4.6.2.b) is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, 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 , b.

With the required containment atmosphere radioactivity monitor inoperable <u, operations may continue for up to 30 days provided:

1.

Grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or 2.

A Reactor Coolant System water inventory balance measurement (Specification 4.4.6.2.b) is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, 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 .

(1)

The provisions of Specification 3.0.4 are not applicable.

BEAVER VALLEY - UNIT 2 3/4 4-17 Amendment No. 64

NPF-73 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) c.

With the required containment sump monitor and the containment atmosphere radioactivity monitor inoperable, 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 4.4.6.1 The leakage detection instrumentation shall be demonstrated OPERABLE by:

a.

Performance of a CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitor at the frequencies specified in Table 4.3-3.

b.

Performance of a CHANNEL CALIBRATION of the required containment sump monitor at least once per 18 months.

BEAVER VALLEY - UNIT 2 3/4 4-18 Amendment No.64

NPF-73 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION

==

3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE, b.

1 gpm unidentified LEAKAGE, c.

1 gpm total primary to secondary LEAKAGE through all steam generators, d.

500 gallons per day primary to secondary LEAKAGE through any one steam generator, and e.

10 gpm identified LEAKAGE.

APPLICABILITY:

MODES 1, 2,

3, and 4.

ACTION:

a.

With any pressure boundary LEAKAGE, be in at least HOT l STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System LEAKAGE greater than any one of the above limits, excluding pressure boundary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 4.4.6.2 Reactor Coolant System LEAKAGES shall be demonstrated to be l within each of the above limits by:

a.

Monitoring the following leakage detection instrumentation at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:cu

1. Containment atmosphere gaseous radioactivity monitor.

(1)

Only on leakage detection instrumentation required by LCO 3.4.6.1.

BEAVER VALLEY - UNIT 2 3/4 4-19 Amendment No. 64

NPF-73 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 2.

Containment atmosphere particulate radioactivity monitor.

3.

Containment sump discharge flow monitor.

4.

Containment sump narrow range level monitor.

b.

Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation.(2)

[

I

}

(2)

Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

BEAVER VALLEY - UNIT 2 3/4 4-20 Amendment No.64

t

, NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350*F if one steam generator becomes inoperable due to single failure i

considerations. Below 350*F, decay heat is removed by the RHR system.

l The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the 4

conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to

design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means j

of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a. manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

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

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary to secondary LEAKAGE 500 gallons per day per steam

=

generator).

Cracks having a primary to secondary LEAKAGE less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary to secondary LEAKAGE of 500 gallons per day per steam generator can l readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging or repair will be required of all tubes with imperfections exceeding the plugging or repair limit.

Degraded steam generator tubes may be 4

BEAVER VALLEY - UNIT 2 B 3/4 4-3 Amendment No.64

-,,..v

-,,-,a

---r

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION BACKGROUND GDC 30 of Appendix A to 10 CFR 50 requires means for detecting and, to the extent practical, identifying the source of RCS LEAKAGE.

Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure.

Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump.

The non-ECCS portion of the containment sump used to collect unidentified LEAKAGE is instrumented to alarm due to abnormal increases in water inventory.

The sensitivity is acceptable for detecting increases in unidentified LEAKAGE.

The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE.

An increase in humidity of the containment atmotJhere would indicate release of water vapor to the containment.

Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE.

Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump.

Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem.

Humidity monitors are not required by this LCO.

BEAVER VALLEY - UNIT 2 B 3/4 4-4 Amendment No. 64

. _ ~

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

BACKGROUND (Continued)

)

Air temperature and pressure monitoring methods may also be used i

to infer unidentified LEAKAGE to the containment.

Containment temperature and pressure fluctuate slightly during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment.

The relevance of temperature and pressure measurements are affected by containment free volume and, for i

temperature, detector location.

Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and pressure monitors are not required by this LCo.

j l

APPLICABLE SAFETY ANALYSES l

The need to evaluate the severity of an alarm or an indication is i

important to the operators, and the ability to compare and verify with indientions from other systems is necessary.

Multiple instrument locations are utilized, if needed, to ensure that the transport delay r

time of the leakage from its source to an instrument location yields an acceptable overall response time.

l The safety significance of RCS LEAKAGE varies widely depending on

[

its source, rate, and duration.

Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakage occur detrimental to the safety of the unit and the public.

i t

LCD One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks.

j This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small l

leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.

i i

BEAVER VALLEY - UNIT 2 B 3/4 4-4a Amendment No. 64 l

i l

i

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

]

LCO (Continued)

The LCO is satisfied when monitors of diverse measurement means are available.

Thus, the containment sump monitor, in combination with a gaseous or particulate radioactivity monitor, provides an acceptable minimum.

The containment sump monitor is comprised of the instruments associated with the non-ECCS portion of the containment sump which monitor narrow range level and sump pump discharge flow.

)

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2,

3, and 4,

RCS leakage detection instrumentation is required to be OPERABLE.

In MODE 5 or 6, the temperature is to be less than or. equal to 200*F and pressure is maintained low or at atmospheric pressure.

Since the temperatures and pressures are far lower than those for MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation are much smaller.

Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

ACTIONS A.

With the required containment sump monitor inoperable, no other form of sampling can provide the equivalent information;

however, the containment atmosphere radioactivity monitoring system will provide indications of changes in leakage.

Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, SR 4.4.6.2.b, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage.

Restoration of the required sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure.

This time is acceptable, considering the frequency and adequacy of the RCS water inventory balance required by Required Action "a."

i BEAVER VALLEY - UNIT 2 B 3/4 4-4b Amendment No.64

NPF REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

ACTIONS (Continued)

Required Action "a" is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is allowed when the containment sump monitor is inoperable.

This allowance is provided because other instrumentation is available to monitor RCS leakage.

b.1.

and b.2.

With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with SR.

4.4.6.2.b, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of the required containment atmosphere radioactivity monitors.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage.

The 30 day Completion Time recognizes at least one other form of leakage detection is

)

available.

Required Action "b"

is modified by a Note that indicates that the provisions of LCO 3.0.4 are not applicable.

As a

result, a MODE change is allowed when the gaseous and particulate containment atmosphere radioactivity monitor channel is inoperable.

This allowance is provided because other instrumentation is available to monitor for RCS LEAKAGE.

g.

With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown is required. The plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems.

BEAVER VALLEY - UNIT 2 B 3/4 4-4c Amendment Nob 4

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

SURVEILLANCE REOUIREMENTS (SR)

SR 4.4.6.1.a SR 4.4.6.1.a requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor.

The check gives reasonable confidence that the channel is operating properly.

l The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 4.4.6.1.a requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitor.

The test ensures that the monitor can perform its function in the desired manner. The test. verifies the alarm setpoint and relative accuracy of the instrument string.

The Frequency of 31 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

SR 4.4.6.1.a also requires the performance of a

CHANNEL CALIBRATION on the required containment atmosphere radioactivity monitor.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

SR 4.4.6.1.1;t SR 4.4.6.1.b requires the performance of a CHANNEL CALIBRATION on the required containment sump monitor.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

The Frequency of 18 months is a typical refueling cycle and considers channel reliability.

Again, operating experience has proven that this Frequency is acceptable.

3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS.

Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systens from the RCS.

BEAVER VALLEY - UNIT 2 B 3/4 4-4d Amendment No. 64

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

BACKGROUND (Continued)

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety.

This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE.

Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on i

its source, rate, and duration.

Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from ' the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100 percent leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, i

in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

i APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE.

However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event.

The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1 gpm primary to secondary LEAKAGE as the initial condition.

BEAVER VALLEY - UNIT 2 B 3/4 4-4e Amendment No.64

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued)

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR).

The leakage contaminates the secondary fluid.

The SLB is more limiting for site radiation releases.

The safety analysis for the SLB accident assumes 1 gpm primary to secondary LEAKAGE.

The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits).

LCR RCS operational LEAKAGE shall be limited to:

i a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further i

deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the RCPB.

LEAKAGE past seals and gaskets is not pressure. boundary LEAKAGE.

Should pressure boundary LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant

System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

b.

Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period.

Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

BEAVER VALLEY - UNIT 2 B 3/4 4-4f Amendment No. 64 i

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Continued) c.

Primary to Secondary LEAKAGE throuch All Steam Generators (SGs)

Total primary to secondary LEAKAGE amounting to 1 gpm through all SGs produces acceptable offsite doses in the SLB accident analysis.

Violation of this LCO could exceed the offsite dose limits for this accident. Primary to secondary LEAKAGE must be included in the total allowable limit for identified LEAKAGE.

d.

Primary to Secondarv LEAKAGE throuch Any One SG The 500 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture.

If leaked through may cracks, the cracks are very small, and the above assumption is conservative.

e.

Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of identified LEAKAGE and is well within the capability of the RCS Makeup System.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a

normal function not considered LEAKACE).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1,

2, 3,

and 4,

the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

BEAVER VALLEY - UNIT 2 B 3/4 4-4g Amendment No.64 j

i NPF-73 REACTOR COOLANT SYSTEM i

1 BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

)

APPLICABILITY (Continued)

LCO 3.4.6.2, "RCS Pressure ~ solation Valve (PIV)," measures leakage through each individual 1~< and can impact this LCO.

Of the l

two PIVs in series in each isolated line, leakage measured through one i

PIV does not result in RCS LEAKAGE when the other is leak tight.

If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

i ACTIONS I

^

A.

If any pressure boundary LEAKAGE exists, the reactor must be i

brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This action I

reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MOD 2 5, the pressure stresses

)

acting on the RCPB are much lower, and further deterioration is much less likely.

l h.

Unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE in excess of the LCO limits must be q

reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This completion 1

Time allows time to verify leakage rates and either identify

)

unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must bo shut down.

This action is i

necessary to prevent further deteriorCcion of the RCPB.

If

)

the unidentified LEAKAGE, identified LEAKAGE, or primary to i

secondary LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its l

potential consequences. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This action reduces the LEAKAGE.

l i

l BEAVER VALLEY - UNIT 2 B 3/4 4-4h Amendment No. 64 i

l NPF-73 t

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS (Continued)

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

SURVEILLANCE REOUIREMENTS (SR)

SR 4.4.6.2 Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

Primary to secondary LEAKAGE is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure.

Therefore, this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring of the leakage detection system is sufficient to provide an early warning of increased RCS LEAKAGE.

These leaksge detection i

systems are specified in LCO 3.4.6.1,

" Leakage Detection Instrumentation."

BEAVER VALLEY - UNIT 2 B 3/4 4-41 Amendment No.64

NPF-73 j

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEIT.T.ANCE REOUIREMENTS (SR) (Continued)

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

Note (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance is required only on leakage detection instrumentation required by LCO 3.4.6.1.

This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to

'be suspended on leakage detection instrumentation which is inoperable or not required to be operable per LCO 3.4.6.1.

Note (2) states that this SR is required to be performed during steady state operation.

3/4.4.6.3 PRESSURE ISOLATION VALVE LEAKAGE The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure.

It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.

Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valve is identified LEAKAGE l and will be considered as a portion of the allowed limit.

j l

BEAVER VALLEY - UNIT 2 B 3/4 4-4j Amendment No.64 l

\\

l i

~

NPF-73 REACTOR COOLANT SYSTEM J

i BASES l

3/4.4.7 CHEMISTRY i

The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the ' structural integrity of the Reactor Coolant System over the life of the plant.

l The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels i

in excess of the Steady State Limits, up to the Transient Limits, for l

the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The j

time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY i

The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately s.3all fraction of 10 CFR Part 100 limits

~

following a steam generator tube rupture accident in conjunction with.

i an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The ACTION statement permitting POWER OPERATION to continue for

)

limited time periods with the primary coolant's specific' activity >

1.0 #Ci/ gram DOSE EQUIVALENT I-131, but within the allowable limit j

shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 yCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval

\\

BEAVER VALLEY - UNIT 2 B 3/4 4-5 Amendment No. 64

NPF-73 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.4 SEAL INJECTION FLOW BACKGROUND The function of the seal injection throttle valves during an accident is similar to the function of the Emergency Core Cooling Systems (ECCS) throttle valves in that each restricts flow from the charging pump header to the Reactor Coolant Systems (RCS).

The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident.

This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during SI.

APPLICABLE SAFETY ANALYSES All ECCS subsysteus are taken credit for in the large break loss of coolant accident (LOCA) at full power.

The LOCA analysis establishes the minimum flow for the ECCS pumps.

The charging pumps are also credited in the small beak LOCA analysis.

This analysis establishes the flow and discharge head at the design point for the charging pumps.

The steam generator tube rupture and main steam line

+

break event analyses also credit the charging pumps, but are not limiting in their design.

Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses.

This LCO ensures that seal injection flow of less than or equal to 28 gpm, with charging pump discharge pressure greater than or aqual to 2410 psig and seal injection flow control valve full open, will be sufficient for RCP seal integrity but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA.

It also ensures that the charging pumps will deliver sufficient water for a

small LOCA and sufficient boron to maintain the core subcritical.

For smaller LOCAs, the charging pumps alone deliver sufficient fluid to overcome the loss and maintain RCS inventory.

BEAVER VALLEY - UNIT 2 B 3/4 5-2 Amendment No.64

NPF-73 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.4 SEAL INJECTION FLOW (Continued) 18d The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that charging pump injection flow is directed to the RCS via the injection points in accordance with the requirements-of 10 CFR 50.46.

The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance.

In order to establish the proper flow line resistance, a pressure and flow must be known.

The flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure and that the charging pump discharge pressure is greater than or equal to the value specified in this LCo.

The charging pump discharge pressure remains essentially constant through all the applicable MODES of this LCO.

A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure.

The valve settings established at the prescribed charging pump discharge pressure result in a conservative valve position should RCS pressure decrease.

The additional modifier of this LCO, the air operated seal injection control valve being full open, is required since the valve is designed to fail open for the accident condition.

With the discharge pressure and control valve position as specified by the LCO, a flow limit is established.

It is this flow limit that is used in the accident analyses.

The limit on seal injection flow, combined with the charging pump discharge pressure limit and an open wide condition of the seal injection flow control valve, must be met to render the ECCS OPERABLE.

If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses.

I i

APPLICABILITY In MODES 1, 2,

and 3, the seal injection flow limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4.

The seal injection flow limit is not applicable for MODE 4 and lower because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES.

Therefore, RCP seal injection flow must be limited in MODES 1, 2,

and 3 to ensure adequate ECCS performance.

BEAVER VALLEY - UNIT 2 B 3/4 5-3 Amendment No.64

o NPF-73 l

EMERGENCY CORE COOLING SYSTEMS BASES

[

f 3/4.5.4 SEAL INJECTION FLOW (Continued) j ACTIONS i

n.

With seal injection flow exceeding its limit, the amount of

(

charging flow available to the RCS may be reduced.. Under i

this Condition, action must be taken to restore the flow to below its limit. The operator has 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the time the l

flow is known to be above the limit to correctly positicn the manual valves and thus be in compliance with the I

accident analysis.

The completion Time minimizes the l

potential exposure of the plant to a LOCA with insufficient injection flow and ensures that seal injection flow is j

restored to or below its limit.

This time is conservative with respect to the Completion Times of other ECCS LCOs; it l

is based on operating experience and is sufficient for taking corrective actions by operations personnel.

l When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be l

initiated.

The completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching i

MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown

rates, and does not challenge plant safety systems or operators.

Continuing the plant shutdown begun in this l

Required Action, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a 5:easonable time,

)

based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEII.TANCE REOUIREMENTS (SR)

SR 3.5.4.1 Verification every 31 days that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained.

The Frequency of 31 days is based on engineering judgment and is consistent with other ECCS valve Surveillance Frequencies.

The Frequency has proven to - be acceptable through operating experience.

As noted, the Surveillance is not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized within a i 20 psig range of normal operating pressure.

The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.

The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the Surveillance is timely.

BEAVER VALLEY - UNIT 2 B 3/4 5-4 Amendment No.64

-.

ML20073G538

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