ML20058E257
| ML20058E257 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/01/1990 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058E261 | List: |
| References | |
| NPF-06-A-110 NUDOCS 9011070098 | |
| Download: ML20058E257 (8) | |
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h, UNITED STATES
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WASHINGTON, D. C. 20555 c.
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ENTERGY OPERATIONS, INC.
DOCKET NO. 50-368 1
i ARKANSAS NUCLEAR ONE, UNIT No. 2 At4ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 110 License No. NPF-6 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Arkansas Power and Light Company (the licensee) dated August 22, 1989, and as supplemented on July 5, i
1990, complies with the standards.and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the i
provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable-assurance:
(1)thattheactivities. authorized 1
'by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be j
conducted in compliance with the Comission's regulations
'D.
The issuance of this licenst amendment will not be inimical to the comon defense and security or to the health and safety of the l
public; and i
E.-
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
I I
9011070098 901101 PDR ADOCK 05000368 P
PNU
i 2
2.
Accordingly, the license is amended by char.ges to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-6 is hereby amended to read as follows:
(2) Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No.110, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the i
Technical Specifications.
3.
The license amendment is effective 30 days 6fter the date of issuance.
FOR THE NUCLEAR REGULATORY C0ftfilSS10N isk N).
Theodore R. Quay, Acting Project Director Project Directorate IV-1 Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of 1ssuance: November 1, 1990
ATTACHMENT TO LICENSE AMENDMENT NO.110
)
FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Anendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE PAGES INSERT PAGES
-3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4 3/4 7-4 3/4 7-4 8 3/4 3-2 B 3/4 3-2 B 3/4 7-1 B 3/4 7-1 1
L
REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIA +1,-3%*.
l APPLICABILITY:
MODES 4 AND 5.
ACTION:
With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE j
1 shutdown cooling loop into operation.
SURVEILLANCE REQUIREMENTS 4.4.2. No additional Surveillance Requirements other than those required by Specification 4.0.5.
"The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
If found outside of a 11% tolerance band, the setting shall be adjusted to within 11% of the lift setting shown.
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k ARKANSAS - UNIT 2 3/4 4-3 Amendment No.110 e
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REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION l
3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift l
setting 2500 psia +1,-3%*.
l APPLICABILITY:
MODES 1, 2 and 3.
ALTION:
a.
With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
The provisions of Specification 3.0.4 may be suspended for one l
valve at a time for up to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> for entry into and during-operation in MODE 3 for the purpose of setting the pressurizer code safety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to heatup.
SURVEILLANCE REQUIREMENTS 4.4.3 No additional Surveillance Requirements other than those required by Specification 4.0.5.
"The lift setting pressure shall correspond to ambient conditions of the I
valve at nominal operating temperature and pressure.
If found outside of a 11% tolerance band, the setting shall be adjusted to within i1% of the lift setting shown.
l i
l l
l ARKANSAS - UNIT 2 3/4 4-4 Amendment No. 36,110
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TABLE 3.7-5 I
35 '
l 9
STEAM LINE SAFETY VALVES l
g; d;
l VALVE NUMBER LIFT SETTING (+1,-3%)*
ORIFICE SIZE
=
l El Line No. 1 Line No. 2 na a.
2.PSV 1002 2 PSV 1052 1078 psig TT (26.0 in.2) b.
2 PSV 1003 2 PSV 1053 1105 psig TT (26.0 in.2)
I c.
2 PSV 1004 2 PSV 1054 1105 psig TT (26.0 in.2) i d.
2 PSV 1005 2 PSV 1055-1132 psig TT (26.0 in.2)
I j
e.
2 PSV 1006 2 PSV 1056 1132 psig TT (26.0 in.2) i t.i.
T
- The lif t setting pressure shall correspond to ambient cor,dition of the valve at
" nominal operating temperature and pressure.
If found outside of a 11% tolerance band, the setting shall be adjusted to within 11% of the lift setting shown.
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3/4.3 INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the mecsurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.
.3/4.3.3.3 SEISMIC INSTRUMENTATION l
The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.
This J
capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100.
The instrumentation is consistent with the recommendations of Safety Guide 12, " Instrumentation for Earthquakes," March, 1971.
L 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION 1
l The OPERABILITY of the meteorological instrumentation ensures that j
sufficient meteorological data is available for estimating potential I
radiation doses to the public as a result of routine or accidental release L
of radioactive materials to the atmosphere.
This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," Feburary 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of' the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY:of the facility from locations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.
l l
ARKANSAS - UNIT 2 B 3/4 3-2 Amendment No. AA, 110 1
r-3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE l'
3/4.7.1.1 SAFETY CYCLES The OPERABILITY of the main steam line code safety valves ensures l
that the secondary system pressure will be limited to within 110% of its j
design pressure of 1000 psig during the most severe anticipated system operational transient.
The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss j
of condenser heat sink (i.e., no steam bypass to the condenser).
1 1
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition.
The "as-found" requirements are conservative l
with respect to Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition, and Addenda through 1987.
The total relieving capacity for l
all valves on all of the steam lines in 14,799,360 lbs/hr which is 118.7 generator ensures that sufficient relieving capacity is available for percent of the total ~ secondary steam flow of 12, 463,439 lbs/hr at 100%
RATED THERMAL POWER.
A minimum of 2 OPERABLE safety valves per steam l
j removing decay heat.
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j STARTUP and/or POWER OPERATION is allowable with safety valves l
inoperable within the limitations of the ACTION requirements on the basis l
of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels.
The reactor trip setpoint reductions are derived on the following bases:
-For two loop operation j
SP = (X) -
) x (125) c For single loop opetation (two reactor coolant pumps operating in the same loop) l l
Sp = (X) - (Y)(U) x (,,)
where:
l SP
=
reduced reactor trip setpoint in percent of RATED l
THERMAL POWER maximum number of inoperable safety valves per steam V
=
line i
l l.
ARKANSAS - UNIT 2 B 3/4 7-1 Amendment No.110 N,,-
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