Text

Proposed Tech Specs Incorporating Administrative Changes to Correct Editorial Errors or Provide Clarification
	ML20215L805
Person / Time
Site:	Beaver Valley
Issue date:	04/29/1987
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20215L438	List: ML20215L435; ML20215L507; ML20215L805;
References
	NUDOCS 8705120379
	Download: ML20215L805 (37)
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- o: ATTACHMENT A Revise the-Technical Specifications as follows:

Remove Page Insert Page 6-13 6-13 6-13a License Condition 2.C(6) ---

B 2-1 B 2-1 2-6 2-6 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 3/4 3 3/4 3-21 3/4 3-27 3/4 3-27 3/4 3-56 3/4 3 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 B 3/4 4-10 B 3/4.4-10 3/4 7-15 3/4 7-15 3/4 9-12 3/4 9-12 3/4 9-13 3/4 9-13 B 2-7 B 2-7 B 3/4 0-1 B 3/4 0-1 B 3/4 0-3 B 3/4 0-3 B 3/4'l-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 5-1 B 3/4 5-1 B 3/4 7 -

B 3/4 7-5 B 3/4 9-3 B 3/4 9-3 d

8705120379 870429

.PDR P

ADOCK 05000334 PDR

ADMINISTRATIVE CONTROLS F

y 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:1

a. The intent of the original procedure is not altered.

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.

c. The change is documented, reviewed by the OSC and approved 1by: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 particulates 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 inhibit- steam generator tube degradation shall be implemented. This 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.

BEAVER VALLEY - UNIT 1 6-13 PROPOSED WORDING

\

m Af>MINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS:

-ROUTINE REPORTS =

6.9.1 In addition to t'.3 applicable reporting requirements of Title l10, ' 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.

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[ -BEAVER VALLEY - UNIT 1 6-13a l

PROPOSED WORDING

~ 2.1 SAFETY LIMITS 4

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

Operation. above the upper boundary of the nucleate boiling regime could result 1 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 percent confidence- level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

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 saturated liquid.

BEAVER VALLEY - UNIT 1 B 2-1 PROPOSED WORDING

TABLE 2.2-1

$ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

m

" FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g 1. Manual Reactor Trip Not Applicable Not Applicable m

k 2. Power Range, Neutron Flux Low Setpoint - 1 25% of RATED Low Setpoint 26% RATED i THERMAL POWER THERMAL POWER C

Z High Setpoint -'i 109% of RATED High Setpoint - 1 110% of a THERMAL POWER RATED THERMAL POWER

~

3. Power Range, Neutron Flux, 1 5% of RATED THERMAL POWER with i 5.5% of RATED THERMAL High Positive Rate a time constant 1 2 seconds POWER with a time constant ~

1 2 seconds

@ 4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 1 5.5% of RATED THERMAL O High Negative Rate a time constant 1 2 seconds POWER with a time constant o 1 2 seconds m

Cy 5. Intermediate Range, 1 25% of RATED THERMAL POWER i 30% of RATED THERMAL gm Neutron Flux POWER m

o 6. Source Range, Neutron Flux i 105 counts per second i 1.3 x 105 counts per z

O second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig

10. Pressurizer Pressure--High 1 2385 psig i 2395 psig

11. Pressurizer Water i 92% of instrument span i 93% of instrument span Level--High

12. Loss of Flow 1 90% of design flow

1 89% of design flow

l per loop per loop-

Design flow is 88,500 gpm per loop.

2

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

-3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting conditions for Operation 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 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 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 inoperable, or solely because its normal power source is inoperable, 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 PROPOSED WORDING

REACTIVITY :'ONTROL 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 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 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.

BEAVER VALLEY - UNIT 1 3/4 1-7 PROPOSED WORDING

REACTIVITY CONTROL SYSTEMS

. CHARGING PUMP SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required 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

c. Verifying pump operation for at least 15 minutes.

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:

a. Verification of an operable RWST pursuant to 4.1.2.7

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-lSI-890C with the plug inserted in its control circuit and an OPERABLE Low l l Head Safety Injection flow path from the RWST to the Reactor Coolant System once per shift, and i 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 .**

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Emergency backup power need not be available l ** 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 PROPOSED WORDING

RkACTIVITY' CONTROL SYSTEMS O

BORIC ACID TRANSFER PUMPS - SHUTDOWN 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 through the boric acid transfer pump of Specification 3.1.2.1.a is l OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no boric acid transfer pump OPERABLE as required to complete the flow path of Specification 3.1.2.1.a, suspend all operations l involving CORE ALTERATIONS or positive reactivity changes until at 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

c. Verifying pump operation for at least 15 minutes.

BEAVER VALLEY - UNIT 1 3/4 1-13 PROPOSED WORDING

hEACTIVITY CONTROL SYSTEMS BORIC ACID TRANSFER PUMPS - OPERATING 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 3.1.2.2.a is OPERABLE. l 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 MARGIN equivalent to 1% ak/k at 200*F; restore at least one boric 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,

b. Verifying, that on recirculation flow, the pump develops a discharge pressure of 1 107 psig, and

c. Verifying pump operation for at least 15 minutes.

BEAVER VALLEY - UNIT 1 3/4 1-14 PROPOSED WORDING

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the + 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 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 + 7%

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

. BEAVER VALLEY - UNIT 1 3/4 2-2 l PROPOSED WORDING

TABLE 3.3-3 (Continued)

. ACTION STATEMENTS

b. Above P-ll or P-12, demonstrate that the Minimum Channels 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 Minimum 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 Total Number of Channels, restore the inoperable channel 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 3.8.1.2, as appropriate, shall apply.

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 a reset of safety injection shall be removed by resetting (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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other 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, or apply Specification 3.0.3.

BEAVER VALLEY - UNIT 1 3/4 3-21 PROPOSED WORDING

TABLE 3.3-5 (Continued)

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

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

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 (Above P-9) l

b. Feedwater Isolation 1 78.0(1)

7. Containment Pressure - Intermediate High-High

a. Steam Line Isolation 1 8.0

8. Steamline Pressure Rate - High Negative

a. Steamline Isolation 1 8.0

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 BEAVER VALLEY - UNIT 1 3/4 3-27 PROPOSED WORDING

TABLE 3.3-5 (C ntinued)

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

4. Steam Line Pressure - Low f a. Safety Injection (ECCS) 1 13.0#/23.0##

l

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

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 (Above P-9) l

b. Feedwater Isolation i 13.0(2)

7. Containment Pressure - Intermediate High-High

a. Steam Line Isolation 1 8.0

8. Steamline Pressure Rate - High Negative

a. Steamline Isolation 1 8.0

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 BEAVER VALLEY - UNIT 1 3/4 3-27 PROPOSED WORDING

. TABLE 3.3-12 (Continued)

ACTION STATEMENTS l 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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months grab samples are analyzed for gross radioactivity (beta or gamma) at a Lower Limit of Detection (LLD) of at least 10" 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.

BEAVER VALLEY - UNIT 1 3/4 3-56 PROPOSED WORDING

TABLE 3.3-13 (Continued)

ACTION STATEMENTS l 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: .

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

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 .

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.

BEAVER VALLEY - UNIT 1 3/4 3-63 PROPOSED WORDING

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 by Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling 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.

l BEAVER VALLEY - UNIT 1 3/4 3-64 PROPOSED WORDING

TABLE 3.3-13 (Continu-d)

-e 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 by Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling 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 .

BEAVER VALLEY - UNIT 1 3/4 3-64 PROPOSED WORDING

REACTOR COOLANT SYSTEM a

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 tc 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.

BEAVER VALLEY - UNIT 1 3/4 4-23 PROPOSED WORDING

1

_~4' TABLE'4.4-5 l_

REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE Vessel Lead Withdrawal Capsule Location Factor Time (EFPY)

V~ 165* 1.37 1 EFPY (Removed)

U 65* .89 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-BEAVER VALLEY - UNIT 1 3/4 4-26 PROPOSED WORDING

REACTOR COOLANT SYSTEM e 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 vessel. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure.

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 assure 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 vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix l H to 10 CFR 50.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pres-sure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are 1 275"F. Either PORV has adequate relieving capability to protect the RCS from over-pressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator 1 25'F above the RCS cold leg temperature or (2) the start of a charging pump and its injection into a water solid RCS.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and opera-tional readiness of these components will be maintained at an accept-able icvel throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

3/4.4.11 RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

BEAVER VALLEY - UNIT 1 B 3/4 4-10 PROPOSED WORDING

PLANT SYSTEMS

. 3/4.7.6 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for all safety related 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 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 , 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. Measurement 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.

DEAVER VALLEY - UNIT 1 3/4 7-15 PROPOSED WORDING

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:

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

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:

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.

BEAVER VALLEY - UNIT 1 3/4 9-12 PROPOSED WORDING

REFUELING OPERATIONS FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.13 The fuel building ventilation system shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool. 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 fuel building ventilation system is restored to OPERABLE status. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 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 fuel storage pool arca at a negativo pressure of

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

c. Testing the SLCRS por Specification 4.7.8.1.

DEAVER VALLEY - UNIT 1 3/4 9-13 PROPOSED WORDING

LIMITING SAFETY SYSTEM SETTINGS BASES reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the foodwater flow by 1 1.55 x 105 lbs/ hour. The Steam l Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess 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 transient on the Reactor Coolant System and steam generators is minimized.

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 voltage or underfrequency to more than one reactor coolant pump. The specified sotpoints 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 undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of 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.

Turbine Trip A Turbino Trip causes a direct reactor trip when operating abovo P-9. Each of the turbine trips provide turbino protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of those trips. Their functional capability at the specified trip sottings is required to enhance the overall reliability of the Reactor Protection System.

i BEAVER VALLEY - UNIT 1 B 2-7 PROPOSED WORDING

l 374n0 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 constituto 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 whose occurrence would violate the intent of the specification. For 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 paramotors as specified in the Limiting conditions for Operation being met without regard for allowable deviations and out of servico provisions contained in the ACTION statements.

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

Exceptions to this provision have boon provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. Those exceptions are stated in the ACTION statements of the appropriate specifications.

DEAVER VALLEY - UNIT 1 B 3/4 0-1 PROPOSED WORDING

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 necessary to ensure the Limiting Conditions for Operation are met and l 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 conditions are provided in the individual Surveillance Requirements.

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, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still 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.

BEAVER VALLEY - UNIT 1 B 3/4 0-3 PROPOSED WORDING

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) 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 principally to the reduction in RCS boron concentration associated with fuel burnup.

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

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 power supply from OPERABLE diesel generators.

With the RCS average temperature above 200*F, a 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 ensure that- minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

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-lSI-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 out of service for maintenance and testing.

BEAVER VALLEY - UNIT 1 B 3/4 1-2 PROPOSED WORDING

P6WER 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 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 bas i ~ are derived from the plant process computer through the AFD Moni'. . 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 l 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.

BEAVER VALLEY -UNIT 1 B 3/4 2-2 PROPOSED WORDING

RIACTOR COOLANT SYSTEM BASES 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 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 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- 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 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 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 case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

BEAVER VALLEY - UNIT 1 B 3/4 4-2a PROPOSED WORDING

ftEACTOR 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 F

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 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 j the accident analyses.

The total steam' generator tube leakage limit of l 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 l Part 100 limits in the event of either a steam generator tube rupture L or steam line break. The 1 gpm limit is consistent with the i assumptions used in the analysis of these accidents. The 500 gpd l

leakage limit per steam generator ensures that steam generator tube

integrity is maintained in the event of a main steam line rupture or l l under LOCA conditions.

l BEAVER VALLEY - UNIT 1 B 3/4 4-3 PROPOSED WORDING i

I f

l E

_ . _ _ . . . _ m ._. _ - _- . _ _ __ _ _ _ _

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 4 C BASES 3/4.5.1 ACCUMULATORS

.The OPERABILITY of each of the RCS accumulators ensures that a sufficient volume of borated. water will be immediately forced into the reactor core through.each of the. cold legs in the event the.RCS pressure falls below the pressure-of the accumulators. This initial i 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 l

an inoperable accumulator minimizes the time exposure of the plant to a LOCA- event occurring concurrent with -failure of an additional i accumulator which may result in unacceptable peak cladding temperatures.

The RCS accumulators are isolated when RCS pressure is reduced to 1000 + 100 psig to prevent borated water from being injected into the RCS during normal plant cooldown and depressurization conditions and also .to prevent inadvertent overpressurization of the RCS at reduced RCS temperature.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS l The OPERABILITY of two separate and independent ECCS subsystems

' ensures that sufficient emergency core cooling capability.will be available in the event of a LOCA assuming the loss of one subsystem

through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying suffi-cient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double _ ended break of the largest RCS cold leg pipe-downward. In

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

The Surveillance Requirements'provided to ensure OPERABILITY of each i' component ensure that at a minimum, the' assumptions used in'the acci- l dent analyses are met and that subsystem OPERABILITY is maintained.

The limitation for a maximum of one charging pump to be OPERABLE and-

, the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 275*F provides assur- l ance that a mass addition pressure transient can be relieved by the operation of a single PORV.

1 3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS

. ensures that sufficient negative reactivity is injected into the core to- limit any positive increase in reactivity caused by RCS system

cooldown. RCS cooldown can be caused by inadvertent depressuri-

zation, a loss-of-coolant accident or a steam line rupture.

I

! The boron injection tank is required to be isolated when RCS temp-l erature is less than 275"F to prevent a potential overpressuri-zation due to an inadvertant safety injection signal.

BEAVER VALLEY - UNIT 1 B 3/4 5-1

! PROPOSED WORDING

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

PLANT SYSTEMS 6 - 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 for1 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 Tuel Handling Accident (FHA) and from leakage -of LOSS OF COOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containment building, such as Engin-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 SLCRS 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 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 and leakage resulting from a FHA will not exceed 10 CFR 100 limits.

3/4.7.9 SEALED SOURCE CONTAMINATION 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 L requiring leak testing, including alpha emitters, is based on 10 CFR

- 70.39(c) limits for plutonium. Leakage of sources excluded from the l l requirements of this specification represent less than one maximum i 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)

I i Deleted l BEAVER VALLEY - UNIT 1 B 3/4 7-5 i PROPOSED WORDING I

REFUELING OPERATIONS l BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ,

l The restrictions on minimum water level ensure that sufficient water l depth is available to remove 99% of the assumed 10% iodine gap l activity released from the rupture of an irradiated fuel assembly. !

The minimum water depth is consistent with the assumptions of the l accident analysis. l 3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure that l 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 PROPOSED WORDING

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-A-T T A C H M E N T B ProposAd Technical Specification Change-No. 133 No Sionificant Hazard Consideration

.h

. Description ^ of amendment. request: Change Request No. 133 would incorporate: those administrative changes identified during_the NRC-and: . Licensee ~ review meeting for the.BV-2. technical specifications.

These changes correct editorial errors or provide _ clarification for consistency- throughout the technical- specifications. This will eliminate -the need. for many extraneous requirements' in the.BV-2

. technical specifications and. in addition cleanup the BV-1 technical specifications.

' 1. Page 6-13, add Administrative Control Section 6 . 8 .~ 5 to

. incorporate license condition:,2.C(6) for implementation of i the . secondary water chemistry monitoring progrkm. .This_will~

J allow the elimination of license condition 2.C(6) since

identical- program requirements are being incorporated into the Administrative Controls.

2. Page B 2-1, .Section - 2.1.1 Reactor Core, add "R-Grid"

, following. "W-3" to clarify .the- correlation used for DNB

. analysis-(recommended by Westinghouse).

i 3. Page 2-6, Table 2.2-1 item 12, correct an editorial error by.

adding :an

to apply the note (Design flow is 88,500 gpm per-loop).

[ 4. Page. 3/4 0-1, Specification- 3.0.1, correct an editorial-error by capitalizing the'"L" in limiting.

5. . Page 3/4 1-7, Specification 3.1.2.1.a and b, correct an i

editorial error by revising the referenced specifications 3.1.2.7a and 3.1.2.7b to 3.1.2.7.a and 3.1.2.7.b respectively.

6. Page 3/4 1-11, Surveillance Requirement 4.1.2.3.3.c, correct an editorial error by capitalizing " OPERABLE".

7.- Page 3/4 1-13, Specification 3.1.2.5 and associated Action U statement, correct an editorial error by _ revising the referenced specification 3.1.2.la to 3.1.2.1.a. ,

8. Page 3/4- 1-14, Specification 3.1.2.6, correct an editorial error by revising the referenced specification 3.1.2.2a to 7 3.1.2.2.a.

9. Page 3/4 2-2, Specification 3.2.1 Action statement b,

, correct an editorial error by revising the referenced Action i 2.a)1) to Action a.2.a)1).

10. Page 3/4 3-21, Table 3.3-3 Action 18, correct an editorial

, error by capitalizing " COLD SHUTDOWN".

j i

, ATTACHMENT B-(Continued)

.Propozed Trchnical Specification Chnnga No. 133

.Page 2-

11. -Page .3/4'13-27, _ Table 3.3-5 item 6.a, provide clarification of; this~ . function ~ by adding a note."(above P-9)" since.a .

reactor tripLfon ; turbine. trip is -enabled above the P-9 permissive setpoint. '

12. Pages~ 3/4 3-56, Table 3.3-12; 3/4 3-63 and 3/4 3-64, Table '

3.3-13:-- correct.an editorial error by changing the heading

-TABLE NOTATION to ACTION STATEMENTS.

P

13. Pages 3/4 4-23, 3/4 4-26, B.3/4 4-10; correct an editorial error by- changing Table 4.4-3 to 4.4-5 since a' Table 4.4-3

. exists on page 3/4 4-14c.

14. Page. 3/4 7-15, revise surveillance requirement (. 7.6.1.b to clarify a commitment setforth in the FSAR SER. :The NRC requested that we add a Flood Alert description to the

, technical specification Bases, however, it has been determined that the surveillance requirement would be a better place to address this concern and that Flood Watch' would differentiate from-the Emergency Plan Alert condition, j- 15. Page 3/4 9-12,. Specification 3.9.12 and Page 3/4 9-13, F Specification 3.9.13 Applicability statements; correct an editorial error by ' changing the statements-to lower case-letters. i

. 16. Page B- 2-7, Steam /Feedwater Flow Mismatch and Low Steam j Generator Water Level, correct ~an editorial error change the

< sign to > 1.55 x 105 lbs/ hour.

l 17. Page B 3/4 0-1, Specification 3.0.4, correct an editorial error change " insure" to " ensure".

18. Page B 3/4 0-3, Specification 4.0.1, correct an editorial-l error change " insure" to " ensure".

. 19. Page B 3/4 1-2, Specification 3/4.1.1.5 Minimum Temperature i for Criticality, correct an editorial error change "NDTT" to j "RTNDT"*

\

L 20. Page B 3/4 2-2, Specification 3/4.2.1 Axial Flux Difference

(AFD), for clarity change "50% & 90% & 15% & 50%" to "50%

and 90% and between 15% and 50%".

21. Page B 3/4 4-2a, Specification 3/4.4.5 Steam Generators, correct an editorial error change " Operating plant" to i " Operating plants".

22. Page B 3/4 4-3, Specification 3/4.4.6.2 Operational Leakage, j correct an editorial error change " rupture as under" to

" rupture or under".

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l

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ATTACHMENT B (Continu:d)

Dropognd Technical Specification Changa No. 133-4 Page.3

~

'23. LPage' B 3/4 5-1, Specification -3/4.5.2 and '3/4.5.3 ECCS Subsystems, correct an editorial error change " ensures" to ,

" ensure". ,

L24.- Page B 3/4 7-5, Specification 3/4.7.9 ' Sealed Source Contamination, correct an editorial error change "10 CFR

'79.39(c)" to "10 CFR 70.39(c)".

25. 'Page B 3/4 3, Specification- 3/4.9.12 and 3/4.9.13 Fuel Building Ventilation System, add additional. clarification of the fuel building area ventilation system and- SLCRS

-functions.

Basis for no significant hazards determination: Based on the criteria- for ' determining whether a significant hazards consideration exists as. setforth in'10 CFR 50.92(c), plant operation in accordance with the proposed amendment would not:

1) Involve a' significant increase in the probability of occurrence or the consequence -of an accident previously evaluated because: These changes are administrative in-nature, intended to correct editorial- errors, clarify specifications and eliminate extraneous . requirements to cleanup the technical specifications. These changes are consistent with current specifications and the UFSAR and do not- reduce or relax present requirements. Therefore, these changes do not involve a significant increase in the probability or consequences of a previously evaluated

-accident. ;

2) Create the possibility of a -new or different kind of accident from any accident previously' evaluated because: No change in plant operations or to equipment or components is

. required. The changes are administrative in nature and will not affect the safe operation of the plant.

3) Involve a significant reduction in the margin of safety because: The changes are administrative in nature, will not affect the safe operation of the plant and do not affect the bases for any technical specification.

h Conclusion I The proposes changes correct editorial errors or provide consistency throughout the technical specifications and will not 1

affect the safety of the plant. Therefore, based on the above, it is proposed to characterize the change as involving no significant hazards consideration.

i t.

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ML20215L805

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