ML20154N475
| ML20154N475 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/20/1988 |
| From: | Black S NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20154N478 | List: |
| References | |
| NUDOCS 8809290318 | |
| Download: ML20154N475 (12) | |
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UNITED STATES g"
' 7, NUCLEAR REGULATORY COMMISSION j
,j WWH NGTON. D. C. 20666 S
8 s,..... f TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEOUOYAH NUCLEAR PLANT, UNIT 1 AMENDitENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. OPR-77.
1.
The Nuclear Regulatory Comission (the Cornission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 27, 1987, 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 (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 conoucted in compliance with the Comission'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 Comission's regulations and all applicable requirements have been satisfied.
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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 pa agraph 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,. 84, 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 A C ---
%A Suzanne Black, Assistant Director for Projects TVA Projects Division Office of Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: September 20, 1988 i
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ATTACHMENT TO LICENSE AMENDMENT N0.84 FACILITY OPERATING LICENSE NO. OPR-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 nuraber.and contain marginal lines indicating the area of change.
Overleaf pages* are provided to maintain document completeness.
REliOVE INSERT 3/4 4-la 3/4 4-la B 3/4 4-1 B 3/4 4-1 I
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REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION
- 3. 4.1. 2 At least two of the reactor coolant loops listed below shall be OPERABLE with at least two reactor coolant loops in operation when the Reactor Trip System breakers are closed and at least one reactor coolant loop in operation when the Reactor Trip System breakers are open:*
a.
Reactor Coolant Lo,op A and its associated steam generator and reactor coolant pump, b.
Reactor coolant Loop B and its associated steam generator and reactor coolant pump, c.
Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, d.
Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY:
MODE 3 ACTION:
a.
With less than the above required reactor coolant loops OPERABLE, restore the required loops 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 HOT ! HUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip System breakers.
c.
With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.2.3 The required Reactor Coolant loops shall be verified to be in h
operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
"All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
SEQUOYAH - UNIT 1 3/4 4-la Amendment No. 12 M
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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-tion, and maintain DNBR above 1.30 during all normal operations and anticipated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
Single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, a single reactor coolant loop or residual heat removal (RHR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
In MODE 5, single failure considerations require that two RHR loops be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point.
The relief capacity of a single fety valve is adequate to relieve any over-pressure condition which could occur during shutdown.
In the event that no SEQUOYAH - UNIT 1 B 3/4 4-1 Amendment No. II* S4
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UNITED STATES E"
NUCLEAR REGULATORY COMMISSION e
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e WASHINGTON, D. C. 20065 k.....,o/
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMEN 0HENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. DPR-79 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 27, 1987, 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 Commission; C.
There is reasonable assurance (i) that the activities authorized by this arnendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the com,on 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 been satisfied.
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2.
Accordingly, the license is an> ended 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. OPR-79 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications, contained in Appendices A and B, ac revised through Amendment No. 75, 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 C0FNISS10N (t Cb
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Suzand Black, Assistant Director for Projects TVA Projects Division Office of Special Projects
Attachment:
Changes to the Technical Specifications Date of issuance: September 20, 1988 I
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ATTACHMENT TO LICENSE AMEN 0 MENT N0.75 FACILITY OPERATING LICENSE fl0.'DPR-79
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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.
Overleaf pages* are provided to maintain document cor.ipleteness.
REMOVE INSERT 3/4 4-1 3/4 4-1*
3/4 4-2 3/4 4-2 B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2*
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I 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR 000LANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.
APPLICABILITY:
MODES 1 and 2.*
ACTION:
With less than the above required reactor coolant loops in operation, be in at least HOT STAN0BY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
SUREVEILLANCE REQUIREMENT 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
"See Special Test Exception 3.10.4.
SEQUOYAH - UNIT 2 3/4 4-1
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITIO$ FbR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERA 0LE with at least two reactor coolant loops in operation when the Reactor Trip System breakers are closed and at least one reactor coolant loop in operation when the Reactor Trip System breakers are open:*
a.
Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, b.
Reactor Coolant Loop B 'and its associated steam generator and Reactor Coolant pump, c.
Reactor Coolant Loop C and its associated steam generatar and Reactor Coolant pump, d.
Reactor Coolant Loop D and its associated steam generator and Reactor Coolant pump.
APPLICABILITY:
MODE 3 ACTION:
a.
With less than the above required Reactor Coolant loops OPERABLE, restore the required loops 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 HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within one hour open the Reactor Trip System breakers.
c.
With no Reactor Coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Reactor Coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.2.3 The required Reactor Coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- All Reactor Coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10'F below scturation temperature.
SEQUOYAH - UNIT 2 3/4 4-2 Amendment No. 75
3/4.4 ' REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, two reactor coolant lobps provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
Single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, a single reactor coolant loop or residual heat removal (RHR) loop prov' ides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
In MODE 5 single failure considerations require that two RHR loops be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
SEQUOYAH - UNIT 2 B 3/4 4-1 Amendment No. 75
REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point.
The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
The power operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the P0,RVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable.
3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR.
The limit is consistent with the initial SAR assumptions.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.
The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
The requirement that 150 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that the plant will be able to control reactor coolant pressure and establish natural circulation conditions.
SEQUOYAH - UNIT 2 B 3/4 4-2 s