Text

Amend 112 to License DPR-66,correcting Errors & Restating Some Specs in Same Way Stated in Unit 2 Tech Specs
	ML20237G537
Person / Time
Site:	Beaver Valley; (DPR-066)
Issue date:	08/14/1987
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Duquesne Light Co, Ohio Edison Co, Pennsylvania Power & Light Co
Shared Package
ML20237G540	List: ML20237G537; ML20237G547;
References
	DPR-66-A-112 NUDOCS 8708240117
	Download: ML20237G537 (35)
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DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOC-KET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 112 i

License No. DPR-66 1.

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

A.

The application for amendment by Duquesne Light Company, et'a1.

(tha-licensee) dated April 29, 1987, 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 p

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; f

and E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-66 is hereby amended to read as follows:

8708240117 870814 PDR ADOCK 05000334 P

pon

2-(2) Technical Specifications J

The Technical Specifications contained in Appendix A, as revised through Amendment No.112, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license is further amended revising page 6 by deleting Paragraph 2.C(6) since its requirements have been.

reloacted to page 6-13 of the Technical Specifications.

Paragraphs 2.C(7),{8) and (9) have been renumbered 2.C(6), (7) and (8),respectively.

4.

This amendmer.t is effective on issuance, to be implemented no later than 30 days after issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l{ &&

et John F. Stolz, Direc or Project Directorate 1-4 Division of Reactor Projects 1/II

Attachment:

1.

Page 6 of license 2.

Changes to the' Technical Specifications Date of Issuance:

August 14, 1987 I

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. 1 (6) Systems Integrity

[

Duquesne Light Company shall implement a program to reduce leakage from systems outside containment that would or could contain highly l

l radioactive fluids during a serious trensient or accident to as low as l

practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

1 (7)

Iodine Monitoring l

Duquense Light Company shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

(8) Backup Method for Determining Subcooling Margin Duquesne Light Company shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

1.

Training of personnel, and i

2.

Procedures for monitoring.

Amendment No. Jr.f,112 A

i.

ATTACHMENT TO LICENSE AMENDMENT NO. 112 FACILITY OPERATING LICENSE'NO. DPR-66 DOCKET NO. 50-334

-Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and

.contain vertical lines indicating the area of change.

Remove

Insert, 2-6 2-6 B 2-1 B 2-1 B 2-7 B 2-7 3/4 0-1 3/4 0-1 3/4 1-7 3/4 1-7 3/4 1-11 3/4 1-11 3/4 1-13 3/4 1-13 3/4 1-14 3/4 1-14 3/4 2-2 3/4 2-2 4

3/4 3-21 3/4 3-21 3/4 3-27 3/4 3-27 3/4 3-56 3/4 3-56 3/4 3-63 3/4 3-63 3/4 3-64 3/4 3-64 3/4 4-23 3/4 4-23 3/4 4-26 3/4 4-26 3/4 7-15 3/4 7-15 3/4 9-12 3/4 9-12 3/4 9-13 3/4 9-13 8 3/4 0-1 B 3/4 0-1 B 3/4 0-3 8 3/4 0-3 B 3/4 1-2 B 3/4 1-2 B 3/4 2-2 B 3/4 2-2 B 3/4 4-2a B 3/4 4-2a B 3/4 4-3 B 3/4 4-3 B 3/4 4-10 B 3/4 4-10 B 3/4 5-1 B 3/4 5-1 B 3/4 7-5 B 3/4 7-5 B 3/4 9-3 B 3/4 9-3 6-13 6-13 6-13a i

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l

2. l' SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating.of the fuel and possible cladding perforation which would result in the release of. fission products to the reactor coolant.

Overheating of'the fuel cladding is prevented by restricting fuel operation to within the I

nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant I

saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could. result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly-measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.

through the W-3 R-Grid correlation.

The W-3 R-Grid DNB correlation l has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux

ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 l

percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

f The curves of Figures 2.1-1, 2.1-2 and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and, average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of l

saturated liquid.

I J

BEAVER VALLEY - UNIT 1 B 2-1 Amendment No. 112 j

LIMITING SAFETY SYSTEM SETTINGS BASES reliability of the Reactor Protection System.

This trip is redundant the Staam Generator Water Level Low-Low trip.

The Steam /Feedwater to Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by 1.55 x 10' lbs/ hour.

The Steam l

Generator Low Water level portion of the trip is activated when the j

water level drops below 25 percent, as indicated by the narrow range These trip values include sufficient allowance in excess instrument.

of normal operating values to preclude spurious trips but will initiate a

reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of

]

the auxiliary feedwater pumps are reduced and the resulting thermal j

transient on the Reactor Coolant System and steam generators is minimized.

o Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of The 5

voltage or underfrequency to more than one reactor coolant pump.

specified setpoints assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For the delay is set so that the time required for a signal undervoltage, reach the reactor trip breakers following the simultaneous trip of to two or more reactor coolant pump bus circuit breakers shall not exceed 0.9 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.3 seconds.

itrbine Trip Trip causes a

direct reactor trip when operating above j

A Turbine P-9.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident analyses for operation of these trips.

Their functional the specified trip settings is required to enhance the capability at overall reliability of the Reactor Protection System.

Amendment No. % 112 i

BEAVER VALLEY - UNIT 1 B 2_7

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY l

LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting conditions for Operation l

l contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a

specification shall exist when the l

requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.

If the Limiting condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

a. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.

Exceptions to these l

requirements are stated in the individual specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements requirements.

This provision shall not prevent passage through OPERATIONAL MODES as required to comply with. ACTION requirements.

Exceptions to these requirements are stated in the individual specifications.

3.0.5 When a system, subsystem, train, component or device is deter-mined to be inoperable solely because its emergency power source is solely because its normal power source is inoperable, inoperable, or it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s),

train (s),

component (s),

and device (s) are OPERABLE or likewise satisfy the requirements of this specification.

Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , action shall be initiated to place the unit in a

MODE in which the applicable Limiting condition for Operation does not apply, by placing it, as applicable, in:

a. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

This specification is not applicable in MODES 5 or 6.

BEAVER VALLEY - UNIT 1 3/4 0-1 Amencment No.14,112

REACTIVITY CONTROL SYSTEMS' 3/4.1.2 BORATION SYSTEMS FLOW' PATHS - SHUTDOWN LIMITING. CONDITION FOR OPERATION 3.1.2.1 As a

minimum, one of the following boron injection flow paths shall.be OPERABLE:

a.

A.

flow path from the boric acid storage. system.via a boric acid-transfer pump to a charging pump to the Reactor Coolant

' System if only the boric acid storage tank.in Specification 3.1.2.7.a is OPERABLE, or l

b.

The flow path from the refueling water storage tank via a charging pump or a low head safety injection pump (with an open 'RCS vent of greater than or equal to 3.14 square o

inches) to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.7.b is OPERABLE.

l APPLICABILITY:

. MODES 5 and 6

.g ACTION With. none of the 'above flow paths OPERABLE, suspend all operations involving CORE -ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.1' At least one of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.

amendment No.M.112 BEAVER VALLEY - UNIT 1 3/4 1-7 L__

m

___m_

.___m__.__._..____.____________.m.____________

.o

REACTIVITY CONTROL SYSTEMS CHARGING PUMP SHUTDOWN LIMITING CONDITION FOR OPERATION i

3.1.2.3 One charging pump in the boron injection flow path required j

by Specification (3.1.2.1) or Low Head Safety Injection Pump (with an open reactor coolant system vent of greater than or equal to 3.14 square inches) shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY:

MODES 5 and 6 ACTION:

With none of the above pumps

OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one

]

charging pump or Low Head Safety Injection pump is restored to

]

OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE at least once per 31 days by:

a.

Starting (unless already operating) the pump from the control room, b.

Verifying, that on recirculation flow, the pump develops a discharge pressure of 1 2402 psig, and Verifying pump operation for at least 15 minutes.

c.

4.1.2.3.2 All charging

pumps, except the above required charging

pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control switches are placed in the PULL-TO-LOCK position and tagged.

4.1.2.3.3 When the Low Head Safety Injection pump is used in lieu of a

charging

pump, the Low Head Safety Injection pump shall be demonstrated OPERABLE by:

Verification of an operable RWST pursuant to 4.1.2.7 a.

b.

Verification of an operable Low Head Safety Injection Pump pursuant to Specification 4.5.2.b.2, c.

Verification of power available*

to MOV-1SI-890C with the plug inserted in its control circuit and an OPERABLE Low l

Head Safety Injecticn flow path from the RWST to the Reactor Coolant System once per shift, and l

d.

Verification that the vent is open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .**

Emergency backup power need not be available Except when the vent path is provided with a valve which is locked or provided with remote position indication, or sealed, or otherwise secured in the open position, then verify these valves open at least once per 7 days.

BEAVER VALLEY - UNIT 1 3/4 1-11 Amendment No '96,710,112

' REACTIVITY CONTROL' SYSTEMS BORIC ACID TRANSFER PUMPS - SHUTDOWN i

LIMITING CONDITION FOR OPERATION 3.1.2.5 One boric acid transfer pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the-flow path l

through the boric acid transfer pump of Specification 3.1.2.1.a is OPERABLE.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With no boric acid transfer pump OPERABLE as required to complete the l

flow path of Specification 3.1.2.1.a, suspend all operations positive reactivity changes until at involving CORE ALTERATIONS or o

least one boric acid transfer pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required boric acid transfer pump shall be demonstrated OPERABLE at least once per 7 days by:

a.

Starting (unless already operating) the pump from the Control room, b.

Verifying, that on recirculation flow, the pump develops a discharge pressure of 1 107 psig, and Verifying pump operation for at least 15 minutes.

c.

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

312

~

REACTIVITY CONTROL SYSTEMS BORIC ACID TRANSFER PUMPS - OPERATING I

LIMITING CONDITION FOR OPERATION 3.1.2.6 At least one boric acid transfer pump in the boron injection flow path required by Specification 3.1.2.2.a shall be l

OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump in Specification l

3.1.2.2.a is OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With no boric acid transfer pump OPERABLE, restore at least one boric acid transfer pump to OPERABLE STATUS within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 borated to a SHUTDOWN o

equivalent to 1%

Ak/k at 200'F; restore at least one boric MARGIN acid transfer pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.6 The above required boric acid pump shall be demonstrated OPERABLE at least once per 7 days by:

a.

Starting (unless already operating) the pump from the Control room, recirculation flow, the pump develops a b.

Verifying, that on discharge prgyvure of 1 107 psig, and Verifying pump operation for at least 15 minutes.

c.

Amendment No. 112 BEAVER VALLEY - UNIT 1 3/4 1-14

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) l l

b.

THERMAL POWER shall not be increased above 90% of RATED l

THERMAL POWER unless the indicated AFD is within the 4 7%

target band and ACTION a.2.a) 1), above has been satisfied.

l c.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the 1

7%

target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED l

o THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its 1 7%

q target band when at least 2 of 4 or 2 of 3 OPERABLE excore channels I

are indicating the AFD to be outside the target band.

POWER

(

OPERATION outside of the 1 7% target band shall be accumulated on a

(

time basis of:

a.

One minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL POWER.

Amendment No. 4.112 BEAVER VALLEY - UNIT 1 3/4 2-2

T ABLE 3. 3 - 3 (continued)

ACTION STATEMENTS b.

Above P-ll or P-12, demonstrate that the Minimum Channels j

OPERABLE requirement is met within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may continue with the inoperable channel bypassed and one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for testing per Specification 4.3.2.1.

ACTION 17 With less than the Mir.imum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed.

ACTION 18 With the number of OPERABLE Channels one less than the j

Total Number of Channels, restore the inoperable channel l

to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 COLD SHUTDOWN within l

the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 33 With the number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kV Bus shall be declared inoperable and the ACTION Statements for Specifications 3.8.1.1 or j

3.8.1.2, as appropriate, shall apply.

l ACTION 34 With the number of OPERABLE Channels one less than the

~

Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until the performance of the next required Channel Functional Test provided the inoperable channel

(

is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 36 The block of the automatic actuation logic introduced by reset of safety injection shall be removed by resetting a

(closure) of the reactor trip breakers within one hour of an inadvertent initiation of safety injection providing that all trip input signals have reset due to stable plant conditions.

Otherwise, the requirements of, action statement 13 shall have been met.

ACTION 37 With the number of OPERABLE channels one less than the Total Number of channels, STARTUP and/or POWER OPERATION

)

may proceed provided the following conditions are

{

satisfied:

a.

The inoperable channel is placed in a

tripped condition within one hour.

b.

The Minimum Channels OPERABLE requirements is met;

however, the inoperable channel may be bypassed for up to 2

hours for surveillance testing of other i

channels per specification 4.3.2.1.1.

ACTION 38 With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window (s)

(bistable status lights or computer checks) that the interlock is in its required state for the existing plant condition, f

or apply Specification 3.0.3.

BEAVER VALLEY - UNIT 1 3/4 3-21 Amendment No. lN,112

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES Initiating Signal and Function Response Time in Seconds t

4.

Steam Line Pressure - Low a.

Safety Injection (ECCS) 1 13.0#/23.0##

b.

Reactor Trip (from SI) 1 3.0 c.

Feedwater Isolation 1 75.0(1) d.

Containment Isolation -

1 22.0#/33.0##

Phase "A" e.

Auxiliary Feedwater Pumps Not Applicable l

f.

Rx Plant River Water 1 77.0#/110.0##

System g.

Steam Line Isolation 1 8.0 5.

Containment Pressure - High-High a.

Containment Quench Spray 1 77.0 b.

Containment Isolation -

Not Applicable Phase "B"

c.

Control Room Ventilation Isolation 1 22.0#/77.0##

6.

Steam Generator Water Level - High-High a.

Turbine Trip-Reactor Trip i 2.5 l

(Above P-9) b.

Feedwater Isolation 1 78.0(1) 7.

Containment Pressure - Intermediate H,igh-High a.

Steam Line Isolation 1 8.0 8.

Steamline Pressure Rate - High Negative

'. Steamline Isolation 1 8.0 a

9.

Loss of Power a.

4.16kv Emergency Bus Under-voltage (Loss of Voltage) 1 1.3 b.

4.16kv and 480v Emergency Bus Undervoltage (Degraded Voltage) 1 95 inenament No. 40,112 BEAVER VALLEY - UNIT 1 3/4 3-27

-w TABLE 3.3-12-(Continued)

ACTION STATEMENTS ACTION'23

- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed provided that prior to initiating a release:

1.-

At 'least two. independent samples are analyzed'in accordance with Specification 4.11.1.1.1, and 2.

At least-two technically qualified members.of the Facility Staff independently verify the release rate calculations and discharge valving;.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 24 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that at least once per 8

hours grab samples are analyzed for gross radioactivity (beta or gamma) at a Lower Limit of Detection (LLD) of at least 10*7 pCi/ml.

ACTION 25 -

With' the number of channels OPERABLE less than required' by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump curves may be used to estimate flow.

ACTION 26 - -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement,. liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank.

Amendment No. TM,112 BEAVER VALLEY - UNIT 1 3/4 3-56

TABLE 3.3-13 (Continued) l ACTION STATEMENTS

' ACTION 27 -

With' the number of channels OPERABLE less'than required by the Minimum channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release:

l 1.

At least two independent samples of the tank's content are analyzed, and j

2.

At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 28 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 ACTION 29 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 30 -

With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, immediately suspend PURGING of Reactor Containment via this pathway if both RM-VS-104A and B

are not operable with the purge / exhaust system in service.

~

i I

BEAVER VALLEY - UNIT 1 3/4 3-63 Amendment No. Y4,112

^ -

L 9

TABLE 3.3-13 (Continued)

ACTION STATEMENTS l

ACTION 31 -

With' the number of channels OPERABLE one: less than required' by the MINIMUM Channels OPERABLE requirement, operation of this system may continue provided grab samples are obtained every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during. additions to.a' tank.

ACTION 32 -

With. the number of' channels OPERABLE less than required L

by Minimum Channels OPERABLE requirement,. effluent releases via this pathway may continue provided samples are continuously collected with auxiliary campling equipment as. required in Table 4.11-2 or sampled and analyzed once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 35 -

See Surveillance 4.11.2.5.1.

~

i l

)

1 Amendment No.14,112 BEAVER VALLEY - UNIT 1 3/4 3-64

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1 a.

The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b.

The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.

c.

The reactor vessel material irradiation surveillance specimens shall be removed and

examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these examinations shall be used l

to update Figures 3.4-2 and 3.4-3.

t l

l Amendment No. 112 BEAVER VALLEY - UNIT 1 3/4 4-23 f

i TABLE 4.4-5 l

REACTOR VESSEL MATERI AL IRRADI ATION SURVEILLANCE SCHEDULE Vessel Lead Withdrawal Capsule Location Factor.

Time (EFPY)'

V 165*

1. 3"!'

1 EFPY (Removed)

U 65*

.f3 3 EFPY W

245'

.89 6 EFPY Y

295*

.89 15 EFPY X.

285*

1.37 EOL T

55'

.58 Standby Z

305*~

.58 Standby:

S 45'

.43 Standby t

BEAVER VALLEY - UNIT 1 3/4 4-26

l l

)

PLANT SYSTEMS i

3/4.7.6 FLOOD PROTECTION y

l I;

LIMITING CONDITION FOR OPERATION

]

l 3.7.6.1 Flood protection shall be provided for all safety related i

systems, components and structures when the water level of the Ohio River exceeds 695 Mean Sea Level at the intake structure.

APPLICABILITY:

At all times.

' ACTION:

With the water level at the intake structure above elevation 695 Mean Sea Level:

a.

Be in at least HOT STANDBY within 6

hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , and b.

Initiate and complete within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the following flood protection measures:

1.

Install and seal the flood doors in the intake structure.

SURVEILLANCE REQUIREMENTS 4.7.6.1 The water level at the intake structure shall be determined to be within the limits by:

a.

Measur7 ment at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the water level is below elevation 690 Mean Sea Level, and b.

Measurement at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , by initiating.a flood watch including communications between plant operators and upstream dam operators, when the water level is equal to or above elevation 690 Mean Sea Level.

l Amendment No. 112 BEAVER VALLEY - UNIT 1 3/4 7-15 1

B REFUELING OPERATIONS FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT LIMITING CONDITION FOR OPERATION 3.9.12 The fuel building ventilation system shall be operating and discharging through at least one train of the SLCRS HEPA filters and

. charcoal-adsorbers during either:

Fuel movement within the spent fuel storage pool, or a.

b.-

Crane operation with loads over the spent fuel storage pool.

APPLICABILITY:

When irradiated fuel which was decayed less than 60.

. days is in the fuel storage pool.

ACTION:

o With.the. requirement of the above specification not satisfied, suspend all operations involving movement of fuel within the storage pool or. crane operation with loads over the storage pool.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS.

4.9.12 The. fuel building ventilation system shall be verified to be operating with all. building doors closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the initiation of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during either fuel movement within the fuel storage pool or crane operation with loads over the fuel storage pool.

Amendment No. 112 BEAVER VALLEY - UNIT 1 3/4 9-12 l

' REFUELING OPERATIONS FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE l

LIMITING CONDITION FOR OPERATION 3.9.13 The fuel building ventilation system shall be OPERABLE.

APPLIC ABILITY:

Whenever irradiated fuel is in the storage pool.

l l

ACTION:

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

The provisions of Specification 3.0.3 are not applicable.

O SURVEILLANCE REQUIREMENTS l

4.9.13 The fuel building ventilation system shall be demonstrated OPERABLE:

a.

At least once per 31 days by initiating flow through the fuel building ventilation system and verifying that the system operates for at least 15 minutes, and b.

At least once per 18 months by:

1.

Verifying that on a

high-high radiation signal, the system automatically directs its exhaust flow through the HEPA filters and charcoal adsorber banks of the Supplemental Leak Collection and Release System (SLCRS).

2.

Verifying that the ventilation system maintains the spent fue'.

storage pool area at a negative pressure of 2

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

c. Testing the SLCRS per Specification 4.7.8.1.

Amendment No, h 112 BEAVER VALLEY - UNIT 1 3/A 13

--_-____ _ _ D

3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Condition for Operation and Surveillance Requirements within Section 3/4.

3.0.1 This specification defines the applicability of each speci-fication in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and would violate the intent of the specification.

For whose occurrence

example, Specification 3.5.1 calls for each Reactor Coolant System accumulator to be OPERABLE and provides explicit ACTION requirements if one accumulator is inoperable.

Under the terms of Specification 3.0.3, if more than one accumulator is inoperable, the unit is required to be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in an least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

As a further example, Specification 3.6.2.1 requires two Containment Spray Systems, to be OPERABLE and provides explicit ACTION requirements if one spray system is inoperable:

Under the terms of Specification 3.0.3, if both of the required Containment Spray Systems are inoperable, the unit is required to be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least COLD SHUTDOWN in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

It is assumed that the unit is brought to the required MODE within the required times by promptly initiating and carrying out the appropriate ACTION statement.

3.0.4 This specification provides that entry into an OPERABLE MODE, or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements.

l The intent of this provision is to ensure that facility operation is initiated with either required equipment or systems inoperable or not other specified limits being exceeded.

this provision have been provided for a limited number Exceptions to of specifications when startup with inoperable equipment would not affect plant safety.

These exceptions are stated in the ACTION statements of the appropriate specifications.

J BEAVER VALLEY - UNIT 1 B 3/4 0-1

,. +_-., o a,nn.

Amendment No.112

{

APPLICABILITY BASES subsystems, trains, components and devices in the other division must be

OPERABLE, or likewise satisfy Specification 3.0.5 (i.e.,

be capable of performing their design functions and have an emergency power source OPERABLE).

In other words, both emergency power sources must be OPERABLE and all redundant

systems, subsystems, trains, components and devices in both divisions must also be OPERABLE.

If these conditions are not satisfied, action is required in accordance with this specification.

In MODES 5

or 6 Specification 3.0.5 is not applicable, and thus the individual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adhered to.

4.0.1 This specification provides that surveillance activities to ensure the Limiting Conditions for Operation are met and l

necessary will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Conditions for Operation are applicable.

Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other provided in the individual Surveillance Requirements.

conditions are Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification.

4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval.

These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations.

The tolerance values, taken either individually or consecutively over 3

test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity.is not significantly degraded beyond that obtained from the nominal specified interval.

4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation.

Under this

criteria, systems or components are assumed to be OPERABLE if the equipment, associated surveillance activities have been satisfactorily performed withi.n the specified time interval.

Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, found or known to be inoperable although still when such items are meeting the Surveillance Requirements.

4.0.4 This specification ensures that the surveillance activities associated with a

Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other applicable condition.

The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a

current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.

Amendment No.1HL 112 BEAVER VALLEY - UNIT 1 B 3/4 0-3

)

____-____A

[

i 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) l l

fuel cycle.

The surveillance requirement for measurement of the MTC at the beginning and near the end of each fuel cycle is adequate to confirm the MTC value since this coefficient changes slowly due i

principally to the reduction in RCS boron concentration associated with fuel burnup.

I 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F.

This limitation is required to ensure (1) the moderator temperature coefficient is within its analyzed temperature

range, (2) the pressurizer is capable of being in an OPERABLE status with a steam I

bubble, (3) the reactor pressure vessel is above its minimum RTUDT o

temperature, and (4) the protective instrumentation is within its normal operating range.

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer

pumps, (5) associated heat tracing

systems, and (6) an emergency 1

power supply from OPERABLE diesel generators.

l With the RCS average temperature above 200*F, e minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders

{

one of the systems inoperable.

Allowable out-of-service periods l

ensure that minor component repair or corrective action may be

{

completed without undue risk to overall facility safety from 1

injection system failures during the repair period.

I With the RCS average temperature less than 200*F, Low Head Safety Injection pump may be used is lieu of the operable charging pump with a minimum open RCS vent of 3.14 square inches.

This will provide latitude for maintenance and ISI examinations on the charging system for repair or corrective action and will ensure that boration and makeup are available when the charging pumps are out-of-service.

An open vent insures that the RCS pressure will not exceed the shutoff head of the Low Head Safety Injection pumps.

MOV-1SI-890C is the Low Head Safety Injection Pump discharge isolation valve to the RCS coldlegs, the valve must be closed prior to reducing RCS pressure below the RWST head pressure to prevent draining into the RCS.

Emergency backup power is not required since this valve is outside containment and can be manually operated if

required, this will allow the associated diesel generator to be taken j

out of service for maintenance and testing.

l l

BEAVER VALLEY - UNIT 1 B 3/4 1-2 Amendment No '9E,112

POWER DISTRIBUTION LIMITS BASES obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect. core burnup considerations.

Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the + 7% target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER Levels.

This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited.

Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty o

deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED -THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50%

and 90% and between 15% and 50% RATED THERMAL POWER, the I

computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B

3/4 2-1 shows a

typical monthly target band near the beginning of core life.

Amencent No.h(,112 BEAVER VALLEY -UNIT 1 B 3/4 2-2

REACTOR COOLANT SYSTEM BASES The Surveillance Requirements for inspection of the steam generator that the structural integrity of this portion of the RCS tubes ensure 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 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 of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expec'ad to be operated in a

manner such that the secondary coolant will be maintained within those parameter limits o

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 500 gallons per day per steam (primary-to-secondary leakage

=

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-l secondary leakage of 500 gallons per day per steam generator can

]

readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown unscheduled inspection, during which the leaking tubes will be and an 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 will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation.

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

laboratory examinations,

tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

teendment No. )A 112 BEAVER VALLEY - UNIT 1 B 3/4 4-2a

REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a

threshold value of less than 1 gpm.

This

~

threshold value is sufficiently low to ensure early detection of o

additional leakage.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 28 gpm with the modulating valve in the supply line fully open at RCS pressures in excess of 2,000 psig.

This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the acci6ent analyses.

The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contri-bution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1

gpm limit is consistent with the assumptions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or I

under LOCA conditions.

Anendment No.112 BEAVER VALLEY - UNIT 1 B 3/4 4-3

4 i

REACTOR COOLANT SYSTEM s

a BASES mm:

/

vessel inside radius are, essentially identical, the measured tran-sition shift for a

sample can be applied with confidence to the adjacent section of the reactor versel.

The heatup and cocidown curves must be recalculated when ttg 4R7,NDT determined frc4 the f

surveillance capsule is different f r6m ' ' the calculated ARTgg7j or the equivalent capsule radiatior expo,sure.

The pressure-temperature limit > lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to sssure compliance with the minimum temperature requirements of Appendix G.to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vesrel irradiation surveillance apecimens and the frequencies for removing and testing these specimens are provided

~

in Table 4.4-5 to assure compliance with the requironents.cf Appendix l

o H to 10 CFR 50.

il The limitations imposed on the pressurizer hcatup and cooldown rates water temperature dif ferential ace. provided to assure that

~

and spray the pressurizer is operated within Cr.e design criteria assumed'for the fatigue

.ine. lysis performed in accordance with the ASMI code t

requirements.

a 1

t The OPERABILITY / df two PORVs or an RCS vent opening of gretter than incLes ensures that the RC.y will be protected from pres-3.14 square transients which could exceed'the limits of Appendi>; G to 10 CFR sure or more of the RCS cold legs are l'2~5'F.

Either Z

. hen one Part 50 w

PORV has adequate relieving capability to protect the RCS.from over-pressurization when the transient is limited to either (1) the sqart, of an idle RCP with the Lsecondary water temperature of the s!.eam generator 1

25'F above the RCS cold leg j temperature or (2? the

,v start of a charci.ng pump and its injection into a wat'er solid RCS.

.I 3/4.4.10 STRUCTUXALINTEGR1TY

)

,1 )

The inservice inspection and testing prograr.: for ASME Code C] ass 1, 2,

and 3

components ensure that the Atructural integrity andfopera-tional readiness ofLthese components will be n,aintained at an accept-able level throughout the life oftths plant.

These programs are in' accordance with Section XI of the ASYd Goiler and Pressure vessel.

Code.and applicable Addenda as requireu; by 10 CFR Part 50.55a(a),

except where specific written relle$

has,,been granted by tho'.;

y Commission pursuant to 10 CFR Part 50.55a(g)16)'1).

3/4.4.11 RELIEF VALVIS The relief valves have remotely operated block valves to provide a

. 3 pcr.itive shutoff capability should a relief valve became inoperable.

The electrical poder for both the 1eliaf valves and th3 block valvea is capable of being supplied fron,sa emergency power caarge to ensure the ability to seal this possible ROE leakage path.

BEAVER VALLEY - UNIT 1

'B 3/4 4-16 Amendment No.14,112

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EMERGENCY CORE COOLING SYSTEMS (ECCS)

ASES 7y 3/4.5.i' ACCUMULATORS The OPERABILITY of each of the RCS accumulators ensures that a sufficient volume of borated water will be immediately forced into through each of the cold legs in the event the RCS the reactor core pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the accident, analysis are met.

The limit of one hour for operation with inoperable accumulator minimizes the time exposure of the plant to an a

LOCA eMent occurring concurrent with failure of an additional accumulator, yhich may result in unacceptable peak cladding temperatures 7

)

The RCS 6/cumulators are isolated when RCS pressure is reduced to y

1000 i 100 psig to prevent borated water from being injected into the ICCf during notgal plant cooldown and depressurization conditions and also fto prevent inadvertent overpressurization of the RCS at reduced

, RCS temperature; J'

/

.2and3/4Mf3'ESCSSUBSYSTEMS r

se OPERABILIIKf of two separate and independent ECCS subsystems iasures that sut'ficient emargency core cooling capability will be

", vailabic in the event of a LOCA assuming the: loss of one subsystem i

/

through anU single failure consideration.

Eitner subsystem operating in conjunction with the accumulators is. capable of supplying suffi-r/

cient core cooling to limit the peak cladding temperatures within

~

acceptable limits for all postulated break sizes ranging from the r

' double ended break of the largest RCS cold leg pipe downward.

In

addition, each ECCS subsystem provides long term core cooling capa-hility inLthe recirculation mode during the accident recovery period.

,r 11he Surveillance Requirements provided to ensure OPERABILITY of each ensure that at a minimum, the assumptions used in the acci-l

,(

component dent analyses are Det and that subsystem OPERABILITY is maintained.

The limitattor for a maximum of one charging pump to be OPERABLE and the Sgtveillance Requirement to verify all charging pumps except the requir ed OPERABLE punp to be inoperable below 275'F provides assur-mass addition pressure transient can be relieved by the ance ;(hat a

.h operction of a single PORV.

3/4."n i.) @ N INJECTION SYSTEM 4

', C The C9E% ABILITY of the boron injection system as part of the ECCS ensuras' that sufficient negative reactivity is injected into che core to 'llmii/ any positive increase in reactivity caused by RCS system coc3d 3wh 7.

RCS cooldown can be caused by inadvertent depressuri-zatica, a loss-of-coolant accident or a steam line rupture.

~ithe: boron injection tank is required to be isolated when RCS temp-

,erature is less than 275'F to prevent a potential overpressure-f zetion due to an inadvertent safety injection signal.

J' SMVER VALLEY - UNIT 1 B 3/4 5-1 ut;g, gg

\\

Amendment No.#,112 7

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PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEM The OPERABILITY of the control room ventilation system ensures that (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in conjunction with control room design provisions, is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole

body, or its equivalent.

This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)

The OPERABILITY of the SLCRS provides for the filtering of postulated radioactive effluents resulting from a Fuel Handling Accident (FHA) and from leakage of LOSS OF COOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containment building, such as Engin-5 eered Safeguards Features (ESF) equipment, prior to their release to the environment.

This system also collects potential leakage of LOCA activity from the Reactor Containment building penetrations into the contiguous areas ventilated by the SLCFS except for the Main Steam Valve Room and Emergency Air Lock.

The operation of this system was assumed in calculating the postulated offsite doses in the analysis for a

FHA.

System operation was also assumed in that portion of the i

Design Basis Accident (DBA) LOCA analysis which addressed ESF leakage following the LOCA, however, no credit for SLCRS operation was taken in the DBA LOCA analysis for collection and filtration of Reactor Containment building leakage even though an unquantifiable amount of contiguous area penetration leakage would in fact be collected and filtered.

Based on the results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA j

and leakage resulting from a FHA will not exceed 10 CFR 100 limits.

3/4.7.9 SEALED SOURCE CONTAMINATION

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The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allow-able limits in the event of ingestion or inhalation of the source material.

The limitations on removable contamination for sources requiring leak testing, including aloha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

Leakage of sources excluded from the l

requirements of this specification represent less than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)

Deleted BEAVER VALLEY - UNIT 1 B 3/4 7-5 Inendment No 'N4,112 1

REFUELING OPERATIONS BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that. sufficient water depth is available to remove 99%

of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.

The minimum water depth-is consistent with the assumptions of the accident analysis.

~3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure that all-radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system _and the resulting iodine removal capacity are consistent with the assumptions of the accident analysis.

The spent fuel pool area ventilation system is non-safety related and only recirculates air through the fuel building.

The SLCRS portion of the ventilation system is safety-related and continuously filters the fuel building exhaust air.

'This maintains a

negative pressure in the fuel building.

BEAVER VALLEY - UNIT 1 B 3/4 9-3 Miendment No. )G, 112

ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

The intent of the original procedure is not altered.

a.

b.

The change is approved by two (2) members of the plant management

staff, at least one (1) of whom holds a Senior Reactor Operator's License on the unit affected.

1 c.

The change is documented, reviewed by the OSC and approved i

by the Plant Manager within 14 days of implementation.

6.8.4 A

Post-Accident monitoring program shall be established, implemented, and maintained:

A program which will provide the capability to obtain and analyze reactor

coolant, radioactive iodines and particulate in plant gaseous effluents, and containment atmosphere samples following an accident.

The program shall include the following:

(i)

Training of personnel, (ii)

Procedures for sampling and analysis, and (iii)

Provisions for maintenance of sampling and analysis equipment.

6.8.5 A

program for monitoring of secondary water chemistry to generator tube degradation shall be implemented.

This inhibit steam program shall be described in the station chemistry manual and shall include:

a.

Identification of a

sampling schedule for the critical parameters and control points for these parameters; b.

Identification of the procedures used to measure the values of the critical parameters; c.

Identification for process sampling points; d.

Procedures for the recording and management of data; e.

Procedures defining corrective actions for off control point chemistry conditions; and f.

A procedure identifying:

1) the authority responsible for the interpretation of the data, and 2) the sequence and timing of administrative events required to initiate corrective action.

Amendment No M,W,W,112 DEAVER VALLEY - UNIT 1 6-13

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORTS 6.9.1.1 A

summary report of plant startup and power escalation testing will be submitted following (1) receipt of an operating

license, (2) amendment to the license involving a planned increase in power

level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifi-cations that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a

description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details requested in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test

program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e.,' initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

BEAVER VALLEY - UNIT 1 6-13a Amendment No. 112 l

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Person / Time
ML20237G537
Site:	Beaver Valley (DPR-066)
Issue date:	08/14/1987
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Duquesne Light Co, Ohio Edison Co, Pennsylvania Power & Light Co
Shared Package
ML20237G540	List: ML20237G537 ML20237G547
References
DPR-66-A-112 NUDOCS 8708240117
Download: ML20237G537 (35)
v • d • e

ML20237G537

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