ML20216F987
| ML20216F987 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/05/1997 |
| From: | Edison G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216F992 | List: |
| References | |
| NUDOCS 9709120156 | |
| Download: ML20216F987 (10) | |
Text
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y UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION g
e WASHINGTON. D.C. 20NGCD1 s.,... /
CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.187 License No. DPR-71 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A, The application for amendment filed by Carolina Power & Light Company (the licensee), dated December 4. 1996, complies with the standards and req)uirements 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 (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 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.
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-71 is hereby amended to read as follows:
p PDR
2 l
(2)
Technical Soecifications The Technical Specifications contained in A)pendices A and B. as revised through Amendment No.187. are here)y incorporated in the license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
Gordon Edison. Acting Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 5,1997
ATTACHMENT TO LICENSE AMENDMENT NO.187 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50 325 Replace the followin the enclosed pages. g sages of the Appendix A Technical Specifications with T1e revised areas are indicated by marginal lines.
Remove Paaes Insert Paaes 3/4 1-2 3/4 1-2 B3/4 1-1 B3/4 1-1 8
l REACTIVITY CONTROL SYSTEMS 3/4.1.2' REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity difference between the actual core k,,, and the predicted core k,,, shall not exceed 1% Ak/k.
APPLICABILITY:
CONDITIONS 1 and 2.
ACTION:
With the reactivity different by more than 1% Ak/k:
a.
Perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected, or b.
Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Submit a Special Test Program to the Commission describing the methods to be used to determine the cause and magnitude of the reactivity difference.
. SURVEILLANCE REQUIREMENTS 4.1.2 The core k,,, shall be predicted and compared to the actual core k,,, for I selected operating conditions:
a.
During the first start-up following CORE ALTERATIONS, and b.
At least once per effective full power month during POWER OPERATION.
BRUNSWICK - UNIT 1 3/4 1-2 Amendment No. 187 l
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTOOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subtritical 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.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold xenon-free condition and shall show the core to be subcritical by at least R + 0.38% delta k/k. The value of R in units of
% delta k/k is tLe difference between the calculated value of maximum core reactivity difring the operating cycle and the calculated beginning-of-life core reactivity.
The value of R must be positive or zero and must be determined for each fuel loading cycle.
Satisfaction of this limitation can be best-demonstrated at the time of fuel loading, but the margin must be determined anytime a control rod is incaphble of insertion.
During the SPIRAL RELOAD deviations from the scheduled core loading are permitted in order to achieve the required 3 cps needed to gain SRM operability provided the cold reactivities (zero voids) of the fuel bundles temporarily loaded around the SRMs are individually less than that of the respective oundles scheduled for those locations.
The cold shutdown margin calculation performed for the scheduled core load 1ng bounds the partially loaded core during the SPIRAL RELOAD process.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can best be demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations.
Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SHUTDOWN MARGIN demonstrations in assuring the reactor can be brought safely to cold, suberitical conditions.
A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.
I "During the first startup following CORE ALTERATIONS" implies that the specified surveillance should be performed upon the initial attainment of a high equilibrium power level, preferably of at least 90% of RATED THERMAL POWER,-during the unit startup.
3/4.1.3 CONTROL RODS The specifications of this section ensure that 1) the minimum SHUTDOWN MARGIN is maintained, 2) the control rod insertion times are consistent with those used in the accident analysis, and 3) the BRUNSWICK - UNIT 1 B 3/4 1-1 Amendment No. 187 l
.________o
nrg g
t UNITED STATES l
g "i
NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 306aHo01 CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50 324 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 218 License No. DPR-62 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by Carolina Power & Light Company (the licensee), dated December 4.1996, 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 (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 amen' ment will not be inimical to the common d
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.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license smendment; and paragraph 2.C.(2) of Facility Operating Liccnse No. DPR-62 is hereby amended to read as follows:
-2 (2)
Technical Soecifications The Technical Specifications contained in A>pendices A and B, as revised through Amendment No. 218. are here)y incorporated in the license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W
Gordon Edison. Acting Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes 'o the Technical Specifications Date of Issuance:
September 5,1997
ATTACHMENT TO LICENSE AMENDMENT NO. 218 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following 3 ages of the Appendix A Technical Specifications with the en;1osed pages.
T1e revised areas are indicated by marginal lines.
Remove Paaes Insert Paoes 3/4 1-2 3/4 1-2 I
B3/4 1-1 B3/4 1-1
._ a
' REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIE'S i
LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity difference between the actual core k, and the predicted g
core k,, shall not exceed 1% ok/k.
APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
With the reactivity different by more than 1% a k/k:
a.
Perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected, or b.
Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Submit a Special Test Program to the Commission describing the methods to be used to determine the cause and magnitude of the reactivity difference.
t SURVEILLANCE REQUIREMENTS
)
4.1.2 The core k,,, shall be predicted and compared to the actual core k,,, for i selected operating conditions:
a.-
During the first start-up following CORE ALTERATIONS, and b..
- At least once per effective full power month during POWER OPERATION.
BRUNSWICK - UNIT 2 3/4 1-2 Amendment No. 218 l
l 3 /4.1 - REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN
_ A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical 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.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold xenon-free condition and shall show the core to be subcritical by at least R + 0.38% delta k/k. The value of R in units of
% delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be determined for each fuel loading cycle. Satisfaction of this limitation can be best demonstrated at the time of fuel loading, but the margin must be determined anytime a control rod is incapable of insertion.
During the SPIRAL RELOAD deviations from the scheduled core loading are permitted in order to achieve the required 3 cps needed to gain SRM operability provided the cold reactivities (zero voids) of the fuel bundles temporarily loaded around the SRMs are individually less than that of the respective bundles scheduled for those locations. The cold shutdown margin calculation performed for the scheduled core loading bounds the-partially loaded core during the SPIRAL RELOAD process.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can best be demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations.
Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SHUTDOWN MARGIN demonstrations in assuring the reactor can be brought safely to cold, subcritical conditions. A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.
I "During the first startup following CORE ALTERATIONS" implies that the specified surveillance should be performed upon the initial attainment of a high equilibrium power level, preferably of at least 90% of RATED THERMAL POWER, during the unit startup.
3/4.1,3 CONTROL RODS The specifications of this section ensure that 1) the minimum SHUTDOWN MARGIN is maintained, 2) the control rod insertion times are consistent with those used in the accident analysis, and 3) the BRUNSWICK - UNIT 2 B 3/4 1-1 Amendment No. 218 l
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