Text

Proposed Tech Specs,Incorporating Administrative Changes for Consistency or for Addl Clarification of Current Requirements
	ML20212G910
Person / Time
Site:	Beaver Valley
Issue date:	02/24/1987
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20212F619	List: ML20212F617; ML20212F648; ML20212G910;
References
	NUDOCS 8703050431
	Download: ML20212G910 (21)
	v • d • e

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' POWER' DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. Whgg the F C is less than or equal to the FR limit xkor the appropriate measured core pYEne, additional power distribution mggs shall be taken and FxC compared to Fky and F xy atleastonceher31EFPD.

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e. The F x limit for Rated Thermal Power (FRTP shall be probided for all core planes containing bank ")D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per specification 6.9.1.14.

f. The F gy limits of e, above, are not applicable in the following core plane regions as measured from the bottom of the fuel:

l 1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

3. Grid plane regions 12% of core height (12.88 inches) measured from grid centerline.

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4. Core plane regions within 12% of core height (12.88 inches) about the bank demand position of the bank "D" control rods.

g. With FC exceeding Fx,L the effects of- F Fo (Z)xkhall be evaluatkd to determine if Fgxy (on Z) is within its limit.

4.2.2.3 When Fo (Z) is measured pursuant to Specification 4.10.2.2, an overall measured Fg (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing

- tolerances and further increased by 5% to account for measurement uncertainty.

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BEAVER VALLEY - UNIT 1 3/4 2-6a PROPOSED WORDING 8703050431 870224 PDR ADOCK 05000334 P PDR

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, ' TABLE-3.3 -

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$ REACTOR TRIP SYSTEM INSTRUMENTATION' -[

s MINIMUM

$ TOTAL NO. CHANNELS CHANNELS APPLICABLE .

p FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES- ACTION-M k '

l. Manual Reactor Trip 2 1 2 1, 2, 3*, 4*, 12' -

8 and 5*

C

$2. Power Range, Neutron Flux a

s a. High Setpoint 4 ,

2 3 l', 2 2

b. Low setpoint 4 2 3 1(1), 2 2

3. Power Range, Neutron Flux 4 2 3 1, 2 2' y High Positive Rate o

@ w4. Power Range, Neutron Flux 4 2 3 1, 2 2 g} High Negative Rate a

.x T5. Intermediate Range, Neutron 2 1 2 1(1), 2, 3*, 3 jN Flux 4* and 5*

a y 6. Source Range, Neutron Flux Q (Below P-10)

a. Startup 2 'l 2 2(2), 3*, 4*, 4 and 5*

b. Shutdown 2 0 1 3, 4, and 5 5

7. Overtemperature AT Three Loop Operation 3 2 2 1, 2 7 l Two Loop Operation 3 1** 2 1, 2 9

8. Overpower AT Three Loop Operation 3 2 2 1, 2 7 l Two Loop Operation 3 1** 2 1, 2 ~9

9. Pressurizer Pressure-Low 3 2 2 1, 2 7 (Above P-7) t

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. TABLE 3.3-5 (Continued)

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ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 4.- Steam Line Pressure-Low 1

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-Phase "A" 1 22.0#/33.0##

e. Auxiliary Feedwater Pumps Not Applicable

f. Rx Plant River Water System 1 77.0#/110.0##

g. Steam Line Isolation 1 8.0

5. Containment Pressure--High-High

a. Containment Quench Spray 1 77.0

b. Containment Isolation-Phase "B" Not Applicable

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

b. Feedwater Isolation i 13.0(2) l

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 Undervoltage 1 1.3 (Loss of Voltage)

b. 4.16kv and 480v Emergency Bus Under- 1 95 voltage (Degraded Voltage)

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

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TABLE 3.3-5 (Continued)

H TABLE NOTATION

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Diesel generator starting and sequence loading delays included.- ' Response. time . limit includes opening of valves to establish SI~ path and attainment of discharge pressure for centrifugal charging pumps and. Low Head Safety Injection pumps.

Diesel generator starting and sequence loading -delays not included. .Offsite power available. Response time limit-

' includes . opening of valves to establish SI path and attainment of discharge pressure for centrifugal ~ charging pumps.

1. Diesel generator starting and sequence loading delays included. Response time. limit includes opening of-valves to establish . RSI -path and attainment of discharge pressure for centrifugal charging pumps.

(1) .Feedwater system overall response time 'shall. include

verification of valve stroke' times applicable to'the feedwater P valves shown for penetrations 76, 77 and 78 on Table 3.6-1.

(2) The 13.0 second response time includes 3 seconds for signal processing -and 10- seconds .for feedwater. flow control valve stroke / closing time (see Table 3.6-l'FCV-lFW-478, 488 and 498).

'#W BEAVER VALLEY -' UNIT 1 3/4 3-28 PROPOSED WORDING D m ' 4 ,.v -

Wg, w , , , - , - ,~~y- g -----w- --+y+,

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TABLE 3.3-6 (Continund)

TABLE NOTATION ACTION 19 -

With the number of channels OPERABLE less than required L by the Minimum Channels OPERABLE requirement, perform-area ' surveys of the monitored area _with portable

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monitoring instrumentation at~ lease once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

r ACTION 20 -

With the number of channels OPERABLE less than required:

by. the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 21 -

With .the number of channels OPERABLE less than required.

by the Minimum Channels OPERABLE requirement, comply with' the applicable ACTION requirements of Specifications 3.9.12 and 3.9.13.

, ACTION'22 - With the number of channels OPERABLE less than required by_ the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

! ACTION 36 - With- the number of OPERABLE channels less than. required L

by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1) Initiate the preplanned _ alternate method of monitoring the appropriate-parameter (s), and

2) Return the channel to OPERABLE status within 30 days, or, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.

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BEAVER VALLEY - UNIT 1 3/4 3-35 PROPOSED WORDING

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'- -INSTRUMENTATION-

~ SEISMIC INSTRUMENTATION LIMITING CONDITION FOR. OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a

a. With the number of OPERABLE seismic monitoring instruments less than required by Table 3.3-7, restore the inoperable instrument (s) to OPERABLE status within 30 days.

b. With one or more seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report ;

to the commission pursuant to Specification 6.9.2 within the next' 10 days outlining the cause of the malfunction.and the plans for restoring the instrument (s) to OPERABLE status.

c. The provisions of Specification 3.0.3 and 3.0.4 are not l applicable.

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SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be ,

4 demcnstrated OPERABLE by the performance of the CHANNEL CHECK,

CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at'the _

frequencies shown in Table 4.3-4.

4.3.3.3.2 A seismic event greater than or equal to 0.Olg shall be reported to the commission within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Each of the above seismic .

< monitoring instruments actuated during a seismic event greater than or equal to 0.0lg shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 30 days following

'the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory $ ground motion. A Special Report shall be prepared and submitted to the

Commission pursuant to Specification 6.9.2 within 30 days describing l the magnitude, frequency spectrum and resultant effect upon facility

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features important to safety.

f BEAVER VALLEY - UNIT 1 3/4 3-38 PROPOSED WORDING

. ~ . _ , . , _ . . ,.--.-..m.-m, , , , . _ , . . - , - --.m. _ . - - . - . . - - - , --, -,,--,.-.,e,,----,.,--w.w

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to TABLE 4 3-6 to

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< REMOTE SHUTDOWN MONITORING INSTRUMENTATION

$ SURVEILLANCE REQUIREMENTS.

E , CHANNEL CHANNEL Q INSTRUMENT '

CHECK CALIBRATION l @ 1. Intermediate Range Nuclear Flux M N.A.

1 m

2. Intermediate Range Startup Rate M N.A.

I I

3. Source Range Nuclear Flux (1) M(4) N.A. l

4. Source Range Startup Rate (1) M(4) N.A. -l m 5. Reactor Coolant Temperature - Hot Leg M R g( .n-

6. Reactor Coolant Temperature - Cold Leg M R.

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@w 7. Pressurizer Pressure M R

$ 8. Pressurizer Level M R to O

.g 9. Steam Generator Pressure M R O

10. Steam Generator Level M R

11. RHR Temperature - HX Outlet (3) M(5) R l

12. Auxiliary Feedwater Flow Rate S/U(2) R Notation (1) Operability required in accordance with specification 3.3.1.1. l (2) Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.a.9 following an extended plant outage. .

(3) Operability required in accordance with Specification 3.4.1.3.

(4) Below P-6.

(5) Channel check to be performed in conjunction with Surveillance Requirement 4.4.1.3.1.

s 9 EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tava 2 350F LIMITING CONDITION FOR OPERATION 3.5.2 Two separate and independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE low head safety injection pump, and

c. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With 'one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN Within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . I

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days describing the circumstances of the j actuation and the total accumulated actuation cycles to date.

W BEAVER VALLEY - UNIT 1 3/4 5-3 PROPOSED WORDING

EMERGENCY. CORE COOLING SYSTEMS

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ECCS SUBSYSTEMS Tavg <350F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump,#

b. One. OPERABLE Low Head Safety Injection Pump, and

c. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4 ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to -OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days describing the circumstances of the l actuation and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.'2.

4.5.3.2 All charging pumps except the above required OPERABLE pumps,

, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever

! .the temperature of one or more of the non-isolated RCS cold legs is 5 275*F by verifying that the control switches are placed in the PULL-TO-LOCK position and tagged.

A maximum of one centrifugal charging pump shall be OPERABLE

whenever the temperature of one or more of the non-isolated RCS cold legs is 5 275*F.

BEAVER VALLEY - UNIT 1 3/4 5-6 PROPOSED WORDING I

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. ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY within one (1) hour.

b. The Safety Limit violation shall be reported to the Commission within onehourandtotheSeniorManagerNuclearl Operations and to the ORC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the on-Site Safety Committee (OSC). This report shall describe (1) . applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the ORC and the Senior Manager Nuclear Operations l within 30 days of the violation. l 6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced-below:

a. The- applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.

b. Refueling operations,

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

l 15.;8.2 Each procedure and administrative policy of 6.8.1 above and changes of intent thereto, shall be reviewed by the OSC and approved by the Plant Manager, predesignated alternate or a predesignated Manager to whom the Plant Manager has assigned in writing the re-sponsibility for review and approval of specific subjects considered by the committee, as applicable. ,

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. BEAVER VALLEY - UNIT 1 6-12 PROPOSED WORDING a

, ADMIN,ISTRATIVE CONTROLS The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR. 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1. The SKYSHINE Code (available from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.

The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21. In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.

The assessment of radiation doses shall be performed in accordance with the ODCM.

The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month period.

RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The x limit for Rated Thermal Power (FRTP shall be providedF [o the Director of the Regional Off[ce) of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Reactor Systems Branch, Division of PWR Licensing-A, U. S. Nuclear Segulatory Commission, Washington, DC 20555 for all core planes containing bank "D" control rods and all unrodded core planes at least 60' days prior to cycle initial criticality In the event that the limit would be submitted at -some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.

Any. information needed to support FgTP will be by request from the NRC and need not be included in khis report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement (Regional Office) within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

BEAVER VALLEY - UNIT 1 6-22 PROPOSED WORDING

L ADMINISTR'ATIVE CONTROLS s

6.lv.2 The following records shall be retained for the duration of

- _the Facility Operating License:

a. ~ Records. and drawing changes reflecting facility design modifications made to systems and equipment described in the . Final Safety Analysis Report.

b. Records of new irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of radiation exposure for all individuals entering radiation control areas.

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components designed for a limited number of -

transients or cycles.

g. Records of training and qualification for current members of the plant staff.

i h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA-Manual,

j. Records of reviews performed for changes made to procedures

- or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the OSC and the ORC. ,

1. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.

m. Records of analyses required by the Radiological Environmental Monitoring Program.

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BEAVER VALLEY - UNIT 1 6-24 PROPOSED WORDING I

$ TABLE 4.4-2 ".

c; STEAM GENERATOR TUBE INSPECTION-Q

% IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION ,

t*

$ Sample Size Result Action Required Result Action Required. Result Action Required A minimum of C-1 None N/A N/A N/A N/A c: S Tubes per

% S. G. C-2 Plug defective t'ubes C-1 None N/A- N/A-8 and inspect addition-e al 2S tubes in this C-2 Plug defective tubes C-1 None S. G. and inspect addition-al 4S tubes in this C-2 Plug defective tubes S. G.

C-3 Perform action for y C-3 result of first o sample m

O mu C-3 Perform action for N/A N/A

$) C-3 result of first ga . sample

$ C-3 Inspect all tubes in All None N/A N/A O

this S. G., plug de- other fective tubes and in- S.G.s spect 2S tubes in each are C-1 other S. G.

Some Perform action.for N/A N/A Notification to NRC S.G.s C-2 result of second pursuant to Specifica- C-2 but sample tion 6.6. no addi-tional S.Gs are C-3 Addi- Inspect all tubes in tional each S.G. and plug S.G. is defective tubes.

C-3 Notification to NRC pursuant to Specifi-cation 6.6.

S=3k% Where N is the number of steam generators in the unit, and n is the number of' steam generators inspected during an inspection.

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3 EMERGENCNCORECOOLINGSYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume by verifying the baron concentration of the accumulator solution.

c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator control circuit is disconnected by removal of the plug in the lock out jack from the circuit.

I 4.5.1.2 Each accumulator water level and pressure alarm channel shall be demonstrated OPERABLE:

a. At least once per 31 days by the performance of a CHANNEL FUNCTIONAL TEST.

b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.

4.5.1.3 During normal plant cooldown and depressurization, each accumulator discharge isolation valve [MOV-lSI-865A, B and C) shall be verified to be closed and de-energized when RCS pressure is reduced to 1,000 i 100 psig.

4 BEAVER VALLEY - UNIT 1 3/4 5-2

. PROPOSED WORDING i

3/4.6 CONTAINMENT SYSTEMS.

3/4.~6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour.or be in at least HOT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s-and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

~ SURVEILLANCE REQUIREMENTS

=

4.6.1.1 Primary CONTAINMENT INTEGRITY L'aall be demonstrated:

a. At least once per 31 days by verifying that:

1. All penetrations

not capable of being closed by l OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.1,

2. All equipment hatches are closed and sealed,

b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

BEAVER VALLEY - UNIT 1 3/4 6-1 PROPOSED WORDING

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ATTACHMENT B Proposed Technical Specification Change No. 122 No Significant-Hazard Consideration-Description -of amendment requ'est: The proposed. amendment would incorporate various. administrative changes to provide consistency or additional clarification of the'following specifications:

1. Page ,3/4 2-6a, Section 3.2.2, Heat Flux Hot- Channel' Factor surveillance requirement- 4.2.2.2.f has been revised by removing reference to specific grid plane regions. .

2. Page 3/4 3-2, Table 3.3-1 items 7 and 8, the applicable action statement has been revised from 2 to 7.

3. Page 3/4 3-27, Table 3.3-5, the response time for item 6.b Feedwater Isolation due to Steam Generator Water Level--High-High has been changed from 178.0 seconds to 113.0 seconds.

4. Page 3/4 3-28, Table 3.3-5, added note (2) applicable to item 6.b to. describe the 13 second response item.

5. Page 3/4 3-35, Table 3.3-6 Action statement 36 has been' revised to reflect specification 3.3.3.9 Action. statement b.

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6. Page 3/4 3-38, Section 3.3.3.3 Seismic Instrumentation, surveillance requirement 4.3.3.3.2 has been revised to reflect the reporting requirements of 10 CFR 50.72 and 10 CFR 50.73.

7. Page 3/4 3-46, Table 4.3-6 note (1) has been replaced with a new note (1) applicable to the Source Range Instruments to indicate these instruments ~ are only required to be operable when required by specification 3.3.1.1. Note (3) has been added applicable to the RHR temperature instrument 'to indicate these instruments are

. only required to be operable when required by specification 3.4.1.3. Note (4) has been added applicable to the Source Range instruments so that the monthly Channel Check is only performed when the plant is below P-6. Note (5) has been added applicable to the RHR Temperature instrument so that the monthly Channel Check is only performed when the RHR system is operable and the applicable surveillance is performed.

8. Page 3/4 5-3, Section 3.5.2 ECCS Subsystems Tavg >350*F, Action statement b has been revised to reflect the reporting requirements of 10 CFR 50.73 and delete the note

applicable to Action statement a.

9. Page 3/4 5-6, Section 3.5.3 ECCS Subsystems Tavg <350 F, intion statement b has been revised to reflect the reporting requirements of 10 CFR 50.73.

1 #- ATTACHMENT-B (cont.)

J' Propeacd Tcchnical:. Specification;Changa No. 122 LPage 2.

i

10. Page '6-12, Section 6.7.1. Safety Limit Violation,':6.7.1.bcand !

6.7.1.d have .beien revised'to-reflect the reporting requirements l of 10 CFR 50.72 and 10~CFR 50.73 respectively.

l

11. Page 6-22, -Section 6.9.1.14 RadialiPeaking Factor, Limit-Report-has been revised to correct the revised copy distribution.

12. Page 6-24, Section 6.10.2, lifetime record retention has been revised by deleting _ item 1 which references. specification 6.13' and items m and n have been renumbered to 1 and m respectively.-

.s 13.-Page .3/4. .4-10d, Table 4.4-2. Steam Generator. Tube Inspection has .

been revised- to ~ - correct the ' required .NRC notification by reference to specification 6.6.

14. Page- 3/4 5-2, surveillance- requirement 4.5.1.1.d has been deleted.

15. Page 3/4 6-1, surveillance requirement 4.6.1.1.a has been revised by adding a note *.

, Basis for no significant hazards determination: The proposed changes do not involve a significant hazards consideration because operation of Beaver . Valley Power Station, Unit;No. 1 in accordance j with these changes would not:

n (A) Involve a significant increase in the probability of occurrence

of the consequence of an accident or malfunction of equipment '

important to safety previously evaluated.

1. Section 3.2.2: The BV-1 core contains fuel assemblies with

three different fuel rod end plugs._. The variation in end

! plug size acts to change the. location of the fuel assembly grids in relation to the bottom of the core. Thus the second grid.is-located at the following elevation in percent of. core height for the various fuel regions:

i i~ Region 1,2,3,4 Region 5,6 Region 7,8 17.8% 17.9% 18.0%

I The proposed change revised surveillance requirement

4.2.2.2.f to specify a generic grid location in lieu of a specific grid location to allow a more accurate application of the grid plane regions in accordance with the applicable fuel rod end plug used. This would i

-s 3, ]

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' ATTACHMENT.B'(cont.)'

Y LPropond'Tichnical: Specification Changa No'.1122 lPage 3:

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affect 'the current- grid elevation by up to 0.2%'(18.0% -

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' 17.8%) however, this. difference is well~within the allowed tolerance of 12%,-therefore, the fuel _ assembly descriptions in.UFSAR Section 3.2 will not be affected.

-2. Table. 3.3 items 7 - and 8: Action statement 7. replaced

. Action ' statement 2 - applicable to items 7 and-8'to provide consistency with 'the actions required for the other 3 channel. Reactor Trip and ESF . functions. Action 2- is appropriate for. 4 channel functions 'whereas, Action 7 is best . applied; to 3 channel functions. When 1. channel is-inoperable, Action 2 allows continued operation..for an-indefinite period, however, Action 7- limits continued.

operation until the next-required CHANNEL FUNCTIONAL' TEST.

, This is consistent with the_ actions required for other 3 channel- functions in the Beaver Valley and the Standard j Technical _ Specifications (STS). The application of Action 7-is. more restrictive than- Action 2 .and is considered administrative in nature since it is consistent with the philosophy applied to similar_3_ channel functions when one 1: channel is inoperable. This change is consistent with the Reactor . Trip System Instrumentation described in UFSAR l Section 7.2 and does not affect the accident analysis of .

Section 14.

3. Table 3.3-5 item 6.b: Feedwater isolation on high-high steam generator level protects against overfilling the steam generator during an excessive feedwater flow event. The feedwater control valves are required to close within 10 seconds on receipt of a steam generator high-high level signal to achieve feedwater isolation and prevent overfill.

This is consistent with the valve action time listed ~in technical specification Table 3.6-1. This change provides clarification of the feedwater isolation requirements such that- .feedwater isolation, by closure of the feedwater control valves, is required within 13 seconds after reaching the steam generator high-high level setpoints. The 13 seconds includes 3 seconds for signal processing and 10 seconds for valve stroke time. This change reflects the feedwater- control valve isolation time listed in technical specification Table 3.6-1 and does not affect the accident

" analysis of UFSAR Section 14.1.9, Excessive Heat Removal Due to Feedwater system malfunctions.

. 4. Table 3.3-5 note (2): This note was added to reflect the -

above change and is consistent with the feedwater control valve isolation time listed in technical specification Table 1 3.6-1 and does not affect the UFSAR accident analysis of Section 14.

l-

5. Table 3.3-6 Action 36: The reporting requirements have been l changed from a special report submitted within 14 days to inclusion in the next Semi-Annual Effluent Release Report.

d v g -vEs e r

e me g e- v == ean g-s --g e- t - ww- m e r w--r - w-- s wr--- e -m m e ow e ,a w,,s.---cw - wre w = r wr--e m -- r-w w e -eww-w.-+ --+----e,-en w w- e r -v ,--r-,,---~~w-----#---r -v -< -

, xa .a-- --- -a -~ ., _, , . = . , . . a s ..a ansa.

L -

L ATTACHMFET B .(cont. );

.o. : Proposed Tachnical LSpecification Chtnga No.-122-Page~4 In addition, the new action statement requires repair.of the inoperable channel and if the , channel cannot be repaired within-'30 days to describe in detail why it.was not repaired.

in' a . timely manner. This is consistent with the reporting requirements. of Action statement.b.in specifications 3.3.3.9 ,

and 3.3.3.10 and does not affect the UFSAR accident-analysis l of Section-14.

6. Section 3.3.3.3: The reporting requirements followingia seismic event specified . in _ surveillance requirement 4.3.~3.3.2 have been changed to requireLnotification of the commission within 1 . hour following a seismic event' greater-than or equal to 0.01g and to change the period when a ,

special report is required from 14 days to 30 days. These changes are consistent with and reflect the requirements of 10 CFR- 50.72 and 10 CFR 50.73 and'do not affect the UFSAR accident analysis of Section 14.

7. Shutdown Table 4.3-6:- Section 3.3.3.5, Remote Instrumentation is applicable in Modes 1, 2 and 3, however, the . applicable modes specified for the Source Range instruments are listed in Table 3.3-1 as 22, 3*, 4* and 5*

for station startup and Modes 3, 4 and 5 for station shutdown and- Modes 4 and 5 for RHR system specified in Section 3.4.1.3. These changes have been incorporated to

- provide clarification of the Remote Shutdown Monitoring Instrumentation operability and surveillance requirements to satisfy the concerns identified in NRC Inspection Report 86-065 The additional notes are applicable to the Source Range and RHR' Temperature. instrumentation to provide consistency with the governing specifications and associated surveillance requirements and are administrative in nature and do not-affect the:UFSAR.

8. Section 3.5.2: Action statement b has been revised to require submittal of a special report within 30 days in p accordance with 10 CFR 50.73. The note

applicable to Action statement a has been deleted since the specified date i

of applicability has expired. These change are administrative in nature and do not affect the UFSAR.

9. Section 3.5.3: Action statement b has been revised to require submittal of a special report within 30 days in accordance with 10 CFR 50.73. This change is administrative in nature and does not affect the UFSAR.

10. Section 6.7.1: Item b has been revised to require notification of the Commission within one hour following a safety limit violation in accordance with 10 CFR 50.72.

Item d has been revised to require a report to the Commission with 30 days following a safety limit violation

in accordance with 10 CFR 50.73. These changes are

[ administrative in nature and do not affect the UFSAR.

, - . - - . _ . _ _ _ _ =..____ _ _ _ __.,_,,. _ - _ . _ _ _ _ _ . . , _ . _ _ _ _ _ . _ . - _

. -f

- ~ *- " ATTACHMENT B (cont.)

10 1 lPropoDed T chnical~ Specification Changa'No. 122 LPage 5 s

i

11. Section 6.9.1.14: This section has been-revised to provide a copy of the RPFLR to the Chieffof the Reactor Systems L Branch .to? reflect a change in the NRC organization. This.i;s an administrative change to comply with a request to correct.

this item'from the NRC staff and does not affect the UFSAR.

I 12 '. Section 6.10.2: Item 1 has been deleted since it references Section 6.13 which was deleted by Amendment 95. Items m and n have been renumbered to 1 and m respectively toireflect

! - the above- change. These changes are administrative in nature and do not affect the UFSAR.

13. Table 4.4-2: This change corrects the NRC notification and reporting requirements to reflect Amendment. 84. The proposed . change is consistent with- the reporting.

requirements specified in surveillance requirement 4.4.5.5.c and Bases Section 3/4.4.5. This is an administrative change and does not affect the UFSAR accident analysis.

14. Surveillance requirements- 4.5.1.1.d: The requirement to-test the automatic actuation feature of the accumulator isolation valves has been deleted to reflect a change in the Westinghouse approved operating procedure. The accumulators are designed to automatically inject borated water into the RCS cold legs in the event:of a rapid.RCS depressurization.. j As required by the operating procedure, during a planned !

depressurization, the accumulator isolation valves are

' manually' closed after the RCS pressure is reduced below-1000 psig to prevent- inadvertent injection of the accumulator contents into the RCS.

During subsequent startups-the valves are then opened,to place the accumulator in service .

before exceeding 1000 psig. >

Current p'rocedures~ require thb accumulator isolation valves to be opened during-normal operation. . The possibility.for inadvertent closure of the isolation valves is eliminated by disconnecting the power - so the valves remain open during normal operation and the valves require no movement to fulfill their safety function. The test requirement for.the valves to open upon generation of an ESF signal is therefore not necessary since the valves are already open during normal operation. Testing the automatic actuation feature of these isolation valves is then not required since this feature is not used during plant operation above 1000 psig or for any other safety function. The proposed change removes an unnecessary surveillance requirement and will provide consistency with plant operation and does not affect the UFSAR accident analysis or the safety of the plant.

15. Surveillance Requirement 4.6.1.1.a: A note has been added to this specification to reflect the Standard Technical Specification and provides clarification of the containment integrity verification requirements for containment

=~ ~

n. .

f -ATTACHMENT B1(cont.)-

0 1Propo cd'T&chnical Specification'Chnnga No.-122

-P, age:6' penetrations inside. containment. The note specifies that all- ~p enetrations ,except valves, blind flanges, and deactivated automatic valves located in containment that are

-locked,. sealed, or otherwise secured in the closed. position ;

shall be verified closed .during each cold shutdown except-

.that. verification need not be performed.more often than'once 4 per 92 . days. This note allows an exception to the position verification- . requirements of. penetration isolation mechanisms inside containment during plant operation so the '

plant operatorsfare not required to enter-the. containment on a monthly basis to verify these penetrations are closed.

Position verification will be provided when,the-plant goes to cold Shutdown.- This is adequate . verification since i containment entries will not be required when'the plant is operating. Therefore, since. containment entries are not required, the position of these penetration ~ isolation ,

mechanisms cannot be- changed and the safety of the plant will'not be affected.

(B) 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.

These changes are administrative in nature and~do not affect the safe. operation of~ the plant. Therefore, these changes will not create. the possibility .of a new or different kind of accident from those described in the Safety Analysis Report. ,

(C) Involve a significant reduction in the margin of safety because:

The changes are administrative in nature and do not significantly ,

affect- the bases for any technical specification and will not affect the safe operation of the plant.

T 4

. Conclusion -

The proposed changes do not involve a significant increase in the ;

probability or consequences of a previously evaluated accident, do not create the possibility of a new or different kind of accident and do not involve a significant reduction in a margin of safety. The changes are administrative in nature and serve to provide consistency between various specifications and additional clarification of current requirements. Therefore, based on the above, it is proposed i to characterize the change as involving no significant hazards i consideration. !

ML20212G910

Text

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