ML20246L662
| ML20246L662 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/11/1989 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246L669 | List: |
| References | |
| NUDOCS 8905180372 | |
| Download: ML20246L662 (14) | |
Text
a carg' #o UNITED STATES y
,g NUCLEAR REGULATORY COMMISSION E
WASHWGTON, D. C. 20586 o***l Q
o TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 115 4
License No. DPR-77 t
I 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 23, 1989, 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 provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l
D.
The issuance of this amendment will not be inimical to the common l
defense and security or to the health and safety of the public; and
]
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have j
been satisfied.
I l
l i
l l
h h, PDC
_ - _ - _ _ _ _ _ _ _ _ _ _ = _
j 2_
q 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 115, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
- %e C
i Suzanne lack, Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 11, 1989 I
f I
1 l
l 1
1 1
1 ATTACHMENT TO LICENSE AMENDMENT NO. 115 j
FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE INSERT 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6*
B 3/4 7-1 B 3/4 7-1*
B 3/4 7-2 B 3/4 7-2
PL' ANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
Two motor-driven auxiliary feedwater pumps, each capable of being a.
powered from separate shutdown boards, and b.
One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2 In addition to the requirements of Specification 4.0.5 each auxiliary feedwater pump shall be demonstrated OPERABLE by :
a.
Verifying that:
1.
each motor-driven pump develops a differential pressure of greater than or equal to the values indicated below on recirculation flow.
1A-A:
1450 psid.
18-B:
1500 psid.
2.
the steam turbine-driven pump develops a differential pressure of greater than or equal to 1201 psid on recirculation flow l
when the secondary steam supply pressure is greater than 842 psig.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
SEQUOYAH - UNIT 1 3/4 7-5 Amendment No.12,115
PLhNT$YSTEMS
[
SURVEILLANCE REQUIREMENTS (Continued) 3.
at least once per 31 days, each automatic control valve in the l
flow path is OPERABLE whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.
b.
At least once per 18 months during shutdown
- by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction pressure test signal.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal.
c.
At least once per 7 days by verifying that each non-automatic valve in the auxiliary feedwater system flowpath is in its correct position.
CThe provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary Feedwater Pump.
SEQUOYAH - UNIT 1 3/4 7-6 Amendment No. 12,77,114
l 3/4.7 PLANT SYSTEMS
~
BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of the system design pressure 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 of condenser heat sink (i.e., no steam bypass to the condenser).
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 total relieving capacity for all valves on all of the steam lines is 1.9 x 107 lbs/hr at 1170 sig which is 127 percent of the total secondary steam flow of 1.493 x 10 lbs/hr at 100% RATED THERMAL POWER.
A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels.
The reactor trip setpoint reductions are derived on the following bases:
For 4 loop operation 3p = (X) - (Y)(V) x (109)
X For 3 loop operation 3p _ (X) - (Y)(U) x (76)
X Where:
SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER l
SEQUOYAH - UNIT 1 B 3/4 7-1
PLANT SYSTEMS i
BASES V = maximum number of inoperable safety valves per steam line U =
maximum number of inoperable safety valves per operating steam line.
109 Power Range Neutron Flux-High Trip Setpoint for 4 loop
=
operation.
76 =
Maximum percent of RATED THERMAL POWER permissible by P-8 Setpoint for 3 loop operation.
X =
Total relieving capacity of all safety valves per steam line in 1bs/ hour, 4.75 x 106 lbs/ hour at 1170 psig.
Y Maximum relieving capacity of any one safety valve in
=
lbs/ hour, 950,000 lbs/ hour at 1170 psig.
3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of off-site power.
The steam oriven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary feedwater pumps are capable of delivering 440 gpm (total feedwater flow) to the entrance of the steam generttors at steam generator pressures of 1100 psia.
At 1100 psia the open steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow.
A total feedwater flow of 440 gpm at pressures of 1100 psia is sufficient to ensure that adequate feedwater l
flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F where the Residual Heat Removal System may be placed into operation.
The surveillance test values ensure that each pump will provide at least 440 gpm plus pump recirculation flow against a steam generator pressure of 1100 psia.
j 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that suf ficient water is available to maintain the RCS at HOT j
STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere concurrent I
with total loss of off-site power.
The contained water volume limit includes an allowance for water not usuable because of tank discharge line location or other physical characteristics.
l l
SEQUOYAH - UNIT 1 B 3/4 7-2 Amendment No.115
L-9
.s** '%
UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
T I
WASHWGTON, D. C. 20055 uSs...../
l TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE.
Amendment No. 105 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 23, 1989, complies with the standards and requirementsoftheAtomicEnergyActof1954,asamended(theAct),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.-
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Connission; C.
There is reasonable assurance (1):that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of-the public; e.nd E.
The issuance of this amendment is in acccrdance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
= = -
l
' l 2.
Accordingly, the license is amended by changes tc the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendice; A and B, as revised through Amendment No.105, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date cf issuance.
FOR THE NUCLEAR REGULATORY COMMISSION nW
/
Suzanne Mack, Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 11, 1989
1 ATTACHMENT TO LICENSE AMENDMENT NO. 105 j
FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 j
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enc'iosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE INSERT 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6*
B 3/4 7-1 B 3/4 7-1*
B 3/4 7-2 B 3/4 7-2 i
l M
PLANT 3YSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate shutdown boards, and b.
One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY:
Modes 1, 2 and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to rostore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2 In addition to the requirements of Specification 4.0.5 each auxiliary feedwater pump shall be demonstrated OPERABLE by:
a.
Verifying that:
1.
each motor-driven pump develops a differential pressure of greater than or equal to the values indicated below on recirculation flow.
2A-A:
1524 psid 28-B:
1464 psid 2.
the steam turbine-driven pump develops a differential pressure of greater than or equal to 1180 psid on recirculation flow when l
the secondary steam supply pressure is greater than 842 psig.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
i SEQUOYAH - UNIT 2 3/4 7-5 Amendment No. 105 l
l PLANT SYSTEMS
~
SURVEILLANCE REQUIREMENTS (Continued) 3.
at least once per 31 days, each automatic control valve in the l
flow path is OPERABLE whenever the auxiliary feedwater system I
is placed in automatic control or when above 10% of RATED THERMAL POWER.
b.
At least once per 18 months during shutdown
- by:
)
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction pressure test signal.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal.
c.
At least once per 7 days by verifying that each non-automatic valve in the auxiliary feedwater system flowpath is in its correct position.
CThe provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary Feedwater Pump.
SEQUDYAH - UNIT 2 3/4 7-6 Amendment No. 68,104 i
b_._________.____-_._-----_.-.____-__-_--_--_-__-_---_--------_-_--------------------------------------------------------------------------------J
l 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of the system design pressure 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 of condenser heat sink (i.e., no steam bypass to the condenser).
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.
Thepotalrelievingcapacityforallvalveson all of the steam lines is 1.9 x 10 lbs/hr at 1 ofthetotalsecondarysteamflowof1.493x10}70poigwhichis127 percent lbs/hr at 100% RATED THERMAL POWER.
A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels.
The reactor trip setpoint reductions are derived on the following bases:
For 4 loop operation SP = (X) - (Y)(V) x 109 l
X l
For 3 loop operation 3p _ (X) - (Y)(U) x 76 X
4 Where:
SP = Reduced reactor trip setpoint in percent of RATED THERMAL POWER V = Maximum number of inoperable safety valves per steam line U = Maximum number of inoperable safety valves per operating steam line SEQUOYAH - UNIT 2 B 3/4 7-1 i
PLANT SYSTEMS BASES SAFETY VALUES (Continued)
I 109 Power Range Neutron Flux-High Trip Setpoint for 4 loop operation
=
Maximum percent of RATED THERMAL POWER permissible by 76
=
P-8 Setpoint for 3 loop operation.
X Total relieving capacity of all safety valves per steam
=
6 line in lbs/ hour, 4.75 x 10 lbs/hr at 1170 psig Maximumrelievinggapacityofanyonesafetyvalvein Y
=
lbs/ hour, 9.5 x 10 lbs/hr at 1170 psig.
3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions in the event of a total loss of off-site power.
The steam driven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary feedwater pumps are capable of delivering 440 gpm (total feedwater flow) to the entrance of the steam generators at steam generator pressures of 1100 psia.
At 1100 psia the cpen steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow.
of 1100 psia is sufficient to ensure that adequate feedwater flow is availableAtotalfeed to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F where the Residual Heat Removal System may be placed into operation.
The surveillance test values ensure that each pump will provide at least 440 gpm plus pump recirculation flow against a steam generator pressure of 1100 psia.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power.
The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the accident analyses.
SEQUOYAH - UNIT 2 B 3/4 7-2 Amendment No. 105
-