ML20059C413
| ML20059C413 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/26/1993 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059C415 | List: |
| References | |
| NUDOCS 9311010147 | |
| Download: ML20059C413 (12) | |
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WASHINGTON, D.C. 205554K)1 TENNESSEE VALLEY AUTHORITY DOCKET N0, 50-327 I
SE0VOYAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendr.ient No. 171 License No. DPR-77
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1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the
~
licensee) dated June 21, 1993, 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 i
provisions of the Act, and the rules and regulations of the Commission; i
C.
There is reasonable assurance (i) that the activities authorized by 3
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.
9311010147 931026 i
PDR ADOCK 05000327-'
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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.171, 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 within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION
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hLACQQ_
Frederick J. Heb on, Director Project Directorate II-4 Division of Reactor Projects - I/II 1
Office of Nuclear Reactor Regulation R
Attachment:
Changes to the Technical i
Specifications Date of Issuance: October 26, 1993 4
,e ATTACHMENT TO LICENSE AMENDMENT NO. 171 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.
REMOVE INSERI B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 6-21 6-21 1
1 a
l 3/4.1 REACTIVITY CONTROL SYSTEMS BASES I
3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN o
A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, _
and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,y, temperature, and is asso-The most restrictive con-dition occurs at E0L, with T,y, at no load operating ciated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.6% delta k/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting. condition and is consistent with FSAR safety analysis assumptions. With T less than 200*F, the reactivity transients resulting from a postulated steam lly,ne break cooldown are -
minimal and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection.
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the Updated Final Safety Analysis Report (UFSAR) analyses.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.
The most negative MTC, which is equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the UFSAR analyses to nominal operating conditions. These corrections involved: (1) a conversion of the MDC used in the UFSAR safety analyses to its equivalent MTC, based on the rate of change of moderator density with tempera-ture at RATED THERMAL POWER conditions; and (2) subtracting from this value the largest differences in MTC observed between end of life (EOL), all rods with-drawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a signifi-cantly more negative E0L MTC at RATED THERMAL POWER.
These corrections trans-formed the MDC value used in the UFSAR safety analyses into the limiting E0L MTC value. The 300-ppm surveillance limit HTC value represents a conservative MTC value at a core condition of 300-ppm equilibrium boron concentration, and is obtained by making these corrections for burnup and soluble boron to the limiting E0L MTC value.
The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly principally because of the reduction in RCS boron concentration associated with fuel burnup.
~
l SEQUOYAH - UNIT 1 B 3/4 1-1 Amendment No.171
,e REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICAllTY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limita-tion is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and
- 5) the reactor pressure vessel is above its minimum RT,y temperature.
3/4.1.2 B0 RATION SYSTEMS The baron 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 350*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 inoper-able. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 4
i B 3/4 1-2 Amendment No. 155, 157, 171 SEQUOYAH - UNIT 1
e ADMINISTRATIVE CONTROLS HONTHLY REACTOR OPERATING REPORT 1
6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
1.
Moderator Temperature Coefficient BOL and E0L limits and 300 pga surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5.
3.
Control Bank Insertion Limits for Specification 3/4.1.3.6.
4 Axial Flux Difference Limits for Specification 3/4.2.1, 5.
Heat Flux Hot Channel Factor. K(z), and W(z) for Specification 3/4.2.2, and 6.
Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specification 3/4.2.3.
6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previo'.isly reviewed and approved by NRC in:
1.
WCAP 9272 P A. " WESTINGHOUSE RELOAD SAFETY EVALUATION HETHODOLOGY", July 1985 (W Proprietary).
(Hethodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 -Control Bank Insertion Limits 3.2.1 - Axial Flux Difference, 3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 Nuclear Enthalpy Hot Channel Factor.)
2.
WCAP-10216 P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F SURVEILLANCE a
TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).
(Hethodology for Specification 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Fim Hot Channel Factor (W(z) surveillance requirements for F MethodologM.)
a 3.
WCAP 10266 P-A Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION H0 DEL USING BASH CODE", March 1987. (W Proprietary).
(Hethodology for Specification 3.2.2 Heat Flux Hot Channel Factor),
4.
WCAP 13631-P A, " SAFETY EVALUATION SUPPORTING A MORE NEGATIVE E0L H0DERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SEQUOYAH NUCLEAR i
PLANTS." HARCH 1993 (W Proprietary).
(Hethodology for Specification 3.1.1.3 Moderator Temperature i
Coefficient) j SEQUOYAH - UNIT 1 6 21 Amendment Nos. 52. 58, 72. 74 117, 152, 155, 156. 171 i
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E UNITED STATES a
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 2055fM)001 j
TENNESSEE VALLEY AUTHORITY i
DOCKET NO. 50-328 1
SE000YAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 161 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 June 21, 1993, 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.
M 4
. 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-79 is hereby
~1 amended to read as follows-(2) Technical Specifications i
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 161, 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 within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION l
i
}
M n Frederick J. HeMon, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 26, 1993 4
.i i
ATTACHMENT TO ' LICENSE AMENDMENT NO.161 FACILITY OPERATING LICENSE NO. OPR-79 i
{
DOCKET NO. 50-328 i
)
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 INSERT J
B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 1
6-22 6-22 l
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3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN HARGIN A sufficient SHUTDOWN HARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition, SHUTDOWN HARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,,,, ~ The most restrictive condition occurs at EOL with T at no load operating temperature, and is associated with a postulated ster.a line bre'al accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN HARGIN of 1.6% delta k/k is required to control the reactivity transient. Accordingly, the SHUTDOWN hARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,
less than 200*F. the reactivt transients resulting from a postulated steam line b,re,ak cooldown are minimal and a #
'lta k/k SHUTDOWN MARGIN provides adequate protection.
3/4.1.1.3 H00ERATOR TEMPELwt COEFFICIENT (HTC)
The limitations on HTC are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the Updated Final Safety Analysis i
Report (UFSAR) analyses.
The HTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of HTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison, The most negative HTC, which is equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the HDC used in the UFSAR onalyses to nominal operating conditions. These corrections involved: (1) a conversion of the HDC used in the UFSAR safety analyses to its equivalent HTC. based on the rate of change of moderator density with temperature at RATED THERHAL POWER conditions; and (2) subtracting from this value the largest differences in HTC observed between end of life (EOL), all rods withdrawn, RATED THERHAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL HTC at RATED THERHAL POWER. These corrections transformed the HDC value used in the UFSAR safety analyses into the limiting E0L HTC value. The 300 ppm surveillance limit HTC value represents a conservative HTC value at a core condition of 300 ppm equilibrium boron concentration, and is obtained by making these corrections for burnup and snluble boron to the limiting EOL HTC value.
The supe llance requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the HTC remains within its limits since this coefficient changes slowly principally because of the reduction in RCS boron concentration associated with fuel burnup.
SEQUOYAH UNIT 2 B 3/4 1-1 Amendnent No.161
.a BfACTIVITY CONTROL SYSTEMS R
i BASES l
3/4.1.1.4 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 it analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTug temperature.
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 350*F, a minimum of two separate and redundant baron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.
The boration capability of either flow path is sufficient to SEQUOYAH - UNIT 2 B 3/4 1-2 Amendment No.
147, 146, 161
$DMINISTRATIVECONTROLS HONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
1.
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion Limits for Specification 3/4.1.3.6.
4.
Axial Flux Difference Limits for Specification 3/4.2.1, 5.
Heat Flux Hot Channel Factor, K(z), and W(z) for Specification 3/4.2.2, and 6.
Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specification 3/4.2.3.
6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP 9272 P-A, " WESTINGHOUSE RELOAD SAFETY EVALVATION HETHODOLOGY", July 1985 (W Proprietary).
(Methodology for Sracifications 3.1.1.3 Moderator Temperature Coefficient, 3.1.3.5 Shutdown Bank Insertion Limit. 3.1.3.6 -Control Bank Insertion Limits, 3.2.1 Axial Flux Difference 3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Hot Channel Factor.)
2.
WCAP 10216 P A. " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F, SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).
-(Methodology for Specification 3.2.1 Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 Heat Flux Hot Channel Factor (W(z) surveillance requirements for F Methodology).)
n 3.
WCAP 10266 P A Rev. 2 "THE 1981 REVISION OF WESTINGHOUSE EVALUATION H0 DEL USING BASH CODE", March 1987, (W Proprietary).
(Methodology for Specification 3.2.2 Heat Flux Hot Channel Factor).
4.
WCAP.13631 P A, " SAFETY EVALUATION SUPPORTING A HORE NEGATIVE EOL HODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SEQUOYAH NUCLEAR PLANTS." HARCH 1993 (W Proprietary).
(Methodology for Specification 3.1.1.3 Moderator Temperature Coefficient) 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
SEQUOYAH UNIT 2 6 22 Amendment Nos. 44, 50, 64, 66, 107, 134, 146, 161
,-