ML20196C372
| ML20196C372 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/19/1998 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20196C380 | List: |
| References | |
| NUDOCS 9812020041 | |
| Download: ML20196C372 (11) | |
Text
-
o nerug 4
g T
UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION
'f WASHINGTON, D.C. 20666 4 001 o
'%,*****,o#
)
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73A License No. DPR-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amer.dment by Tennessee Valley Authority (the licensee) dated August 21,1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9812O20041 981119 PDR ADOCK 05000327 p
.___m 2.
' Accordingly, the licente 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-79 is hereby amended to read as follows:
l (2) Technical Soecifications i
The Technical Specifications contained in Appendices A and B, as revised through l
Amendment No238. 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, to be implemented no l
later than 45 days afterissuance.
FOR THE NUCLEAR REGULATORY COMMISSION h
Frederick J. Hebdon, Director Project Directorate ll-3 Division of Reactor Projects -l/II l
Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications Date of issuance: tb.* der 19,1998 l-i' 1
-=
l i
I ATTACHMENT TO LICENSE AMENDMENT NO.23R i
FACl,LITY OPERATING LICENSE NO. DPR-77 1
l DOCKET NO. 50-327 i
l
=
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 marginallines indicating the area of change.
REMOVE INSERT 3/4 7-7 3/4 7-7 B 3/4 3-2 B 3/4 3-2 8 3/4 3-3 B 3/4 3-3 9
1
v s
I PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPEP,ATION 3.7.1.3 A condensate storage tank system (CST) shall be OPERABLE with a FP contained water volume of at least 190,t)00 gallons of water.
APPLICABILITY: MODES 1, 2 and 3, MODE 4 when steam generator is relied upon for heat removal.
b_qlI_OJ:
I With the condensate storage tank system inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:
a.
Restore the CST to OPERABLE status 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without reliance on steam generator for heat removal, or l
b.
Verify by administrative means OPERABILITY of the Essential Raw Cooling Water System as a backup supply to the auxiliary feedwater pumps
- and restore the condensate storage tank to OPERABLE status within 7 days 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without reliance on steam generator for heat removal.
9 SURVEILLANCE REQUIREMENTS 4;7.1.3.1 The condensate storage tank system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.
OPERABILITY shall be verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following initial l
verification.
SEQUOYAH - UNIT 1 3/4 7-7 Amendment No.P.%
l I
v s
I INSTRUMENTATION i
BASES l
t l
The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for,those channels with response times indicated as not applicable in the updated final safety analysis report.
l Response time may be demonstrated by any series of sequential, l
overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.
Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing maintenance. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis performed in iR58 1
WCAP-10271, Supplement 1, which determines bypass breaker availability.
l The placing of a channel in the trip condition provides the safety BR-9 function of the channel.
If the channel is tripped for testing and no other condition would have indicated inoperability, the channel should not be declared inoperable.
The Auxiliary Feedwater (AFW) Suction Pressure-Low function must be OPERABLE in MODES 1, 2, and 3 to ensure a saf,ety grade supply of water for the AFW System to maintain the steam generators as the heat sink for the reactor. t This function does not have to be OPERABLE in MODES 5 and 6 because heat being generated in the reactor is removed via the Residual Heat Removal (RHR) System and does not require the steam generators as a heat sink.
In MODE 4, AFW automatic suction transfer does not need to be OPERABLE because RHR will already be in operation, or sufficient time is available to place RER in operation to remove decay heat.
l 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 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.
The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
1 For the purpose of measuring F (X,Y,Z) or F,w (:X,Y) a full incore flux map IR227 n
is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incere thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
SEQUOYAH - UNIT 1 B 3/4 3-2 Amendment No. 54, 190,234 f
+
l i
\\
l INSTRUMENTATION f
BASES 1
3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly det* ermine the magnitude of a seismic event and evaluate the response of those features important to safety.
This l
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 l
required pursuant to Appendix "A" of 10 CFR Part 100.
All specified
- measurement ranges represent the minimum ranges of the instruments. This instrumentation is consistent with the recommendations of Regulatory R85 Guide 1.12, " Instrumentation for Earthquakes," April 1974.
3 /4. 3. 3. 4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release 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 I
public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION j
The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility and the potential capability for subsequent cold shut-BR down 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 Criterion 19 of 10 CFR 50.
3/4.3.3.6 CHLORINE DETECTION SYSTEMS
(
This specification deleted.
R66 3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2,
" Instrumentation R153 for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.
The postaccident monitoring instrumentation limiting condition for operation provides the requirement of Type A and Category 1 monitors that provide information required by the control room operators to:
Permit the operator to take preplanned manual actions to accomplish safe plant shutdown.
l Determine whether systems important to safety are performing their intended functions.
SEQUOYAH - UNIT 1 B 3/4 3-3 Amendment No. M, 91, 1+9, -159, 238 Jel-y-s'V-1932 1
ka %q p
UNITED STATES g
j NUCLEAR REGULATORY COMMISSION sa f
WASHINGTON, D.C. 20666 4 001 4.....,o TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. PPA 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) dat'ed August 21,1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
l l
l i
i l
2 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 Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 228 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, to be implemented no later than 45 days after issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y
Frederick J. Hebdon, Director i
Project Directorate ll-3 Division of Reactor Projects -l/II Office of Nuclear Reactor Regulation Attachments: 1. Changes to the Technical Specifications l
Date of issuance: Noverber 19, 1998 i
I 1
1-a h
l
ATTACHMENT TO LICENSE AMENDMENT NO. ??R FACl,LITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 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.
REMOVE INSEBI 3/4 7-7 3/4 7-7 B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 l
I l
l L
i
O PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank system (CST) shall be OPERABLE with a contained water volume of at least 190,D00 gallons of water.
APPLICABILITY: MODES 1, 2 and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTION:
With the condensate storage tank system inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:
a.
Restore the CST to OPERABLE status 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 SEUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without reliance on steam generator for heat removal, or l
b.
Verify by administrative means OPERABILITY of the essential raw cooling water system as a backup supply to the auxiliary feedwater pumps
- and restore the condensate storage tank to OPERABLE status within 7 days 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without reliance on steam generator for heat removal.
. r' SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The condensate storage tank system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume is within its limits when the system is the supply source for the auxiliary feedwater pumps.
OPERABILITY shall be verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following initial verification.
SEQUOYAH - UNIT 2 3/4 7-7 Amendment No.234
e e
s I
l JNSTRUMENTATION l
l l
BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM l
INSTRUMENTATION (Continued) l The measurement of response time pt the specified frequencies provide =
assurance that the protective and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with I
response times indicated as not applicable in the updated final safety analysis l
report.
R182 Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test l
measurements or 2) utilizing replacement sensors with certified response times.
Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing R46 maintenance.
The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis performed in WCAP-10271, Supplement 1, which determines bypass breaker availability.
The placing of a channel in the trip condition provides the safety functior of the channel.
If the channel is tripped for testing and no other BR-10 conditi<
would have indicated inoperab211ty,,the channel should not be declar'
.noperable.
- te Auxiliary Feedwater (AFW) Suction Pressure-Low function must be OPERAbu in MODES 1, 2,
and 3 to ensure a safety grade supply of water for the AFW System to maintain the steam generators as the heat sink for the reactor.
This function does not have to be OPERABLE in MODES 5 and 6 because heat being generated in the reactor is removed via the Residual Heat Removal (RHR) System and does not require the steam generators as a heat sink.
In MODE 4, AFW automatic suction transfer does not need to be OPERABLE because RER will already be in operation, or sufficient time is available to place RHR in operation to remove decay heat.
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.?
MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.
The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring F (X,Y,Z) or F,g (X,Y) a full incore flux map lR214 n
is used.
Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment Nos.
44,.3.2, 132, 22A
l
,, s
(
INSTRUMENTATION l
BASES 4
3/4.3.3.3 SEISMIC INSTRUMENTATION 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 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.
All specified R72 measurement ranges represent the minimum ranges of the instruments. The instrumentation is consistent with the recommendations of Regulatory Guide 1,12, " Instrumentation for Earthquakes," April 1974.
3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that i
sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release 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," February 1972.
3 /4. 3. 3. 5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that suf-ficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility and the potential capability for subsequent cold shut-down from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General BR Design Criterion 19 of 10 CFR 50.
l 3/4.3.3.6 CHLORINE DETECTION SYSTEMS R54 l
This specification deleted.
3 /4. 3. 3. 7 ACCIDENT MONITORING INSTRUMENTATION l
The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and l
assess these variables following an accident. This capability is consistent l
with the recommendations of Regulatory Guide 1.97, Revision 2,'" Instrumentation R135 l
for Light-Water-Cooled Nuclear Power Plants te Assess Plant Conditions During l
and Following an Accident," December 1980.
l The postaccident monitoring instrumentation limiting condition for l
operation provides the requirement of Type A and Category 1 monitors that R149 provide information required by the control room operators to:
Permit the operator to take preplanned manual action to accomplish safe plant shutdown i
i July.JL.1122 SEQUOYAH - UNIT 2 B 3/4 3-3 Amendment Nos. 45, 46, 44, 7.2, 145, 449,2%
i
- -. - _