ML20245J195
| ML20245J195 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/19/1989 |
| From: | Reeves E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245J201 | List: |
| References | |
| NUDOCS 8905040044 | |
| Download: ML20245J195 (12) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 5
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, WASHING TON, D. C. 20555 i
CAROLINA POWER & LIGHT COMPANY, et al.
D_0CKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AME'.34ENT TO FACILITY OPERATING LICENSE Amendment No.127 License No. DPR-71 1.
The Nuclear Regulatory Connission (the Consission) has found that:
A.
The application for amendment filed by Carolina Power & Light Company (the licensee), dated August 3, 1987, 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; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Consission; C.
There is reasonable assurance: (i)thattheactivitiesauthorized 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 Connission'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:
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- i (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.127, are hereby incorporated in the l
license. Carolina Power & Light Company shall operate the facility 1
in accordance with the Technical Specifications.
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This license amendment is effective as of the date of its issuance and
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shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor Adensam/for I
Edward A. Reeves, Acting Director Project Directorate II-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the.Techa* cal Specifications Date of Issuance:
April 19,1989
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ATTACHMENT TO LICENSE AMENDMENT N0.127 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages 3/4 1-14 3/4 1-14 8 3/4 1-3 8 3/4 1-3 B 3/4 1-4 8 3/4 1-4
REACTTVITY C0KIROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITINC CONDITION FOR OJERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be OPERABLE when THERMAL POWER is less than 20% of RATED THERMAL POWER.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*.
l ACTION:
i i
With the RWM inoperable, the provisions of Specification 3.0.4 are not i
applicable, operation may continue and control rod movement is permitted provided that a second licensed opera:or.or other qualified member of the technical staff is present at the reactor control console and verifies compliance with the prescribed control rod pattern.
SURVEILLANCE REQUIREMENTS 4.1.4.1.1 The RWM shall be demonstrated OPERABLE in OPERATIONAL CONDITION 2, l
prior to withdrawal of control rods for the purpose of making the reactor critical and in OPERATIONAL CONDITION 1 when the RWM is initiated during l
control rod insertion when reducing THERMAL POWER bys A.
Verifying proper annunciation of the selection error of at least one out-of-sequence control rod, and b.
Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
4.1.4.1.2 The RWM shsil be demonstrated OpskABLE by verifying the control rod Banked Position Withdra&al Sequence input to the RWM computer is correct l
following any loading of the sequence program into the computer.
- Entry into OPERATIONAL C0KDITION 2 and withdrawal of selected control rods is l
permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
BRUNSWICK - UNIT 1 3/4 1-L4 Amendment No.127
REACTIVITY CONTROL SYSTEM BASES COWTROL RODS (Continued) ou a scram than has been analyzed even though control rods with inoperable accumulators may stiLL be inserted with normal drive water pressure.
Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactors.
Control rod coupling integri'ty is required to ensure compliance with the analysis of the rod drop accident in the-FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and, therefore, this check must be performed prior to achieving criticality after reach refueling. The subsequent check is performed as a backup to the initial demonscrson.
In order to ensure that the control rod patterns can be followed and, therefore, that other parameters are within their limits, the control rod position indication system must be OPERA 8LE.
The ccntrol rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawai is less than a normal withdrawai increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL E00 PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximus in sequence individual control rod or control rod segments which ara withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than or equal to 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design race of the velocity limiter, could result in a peak enthalpy of 280 cal /ge. Thus, requiring the RSCS and RWM to be Oi:RA8LE'when THERMAL POWER is less than 20% of RATED THERMAL POWER provides adequate control.
Use of the Banked Position Withdrawal Seque' ace (BPWS) ensures that in the event of a control rod drop accident the peak fuel enthalpy will not be greater than 280 cal /sm (Reference 4).
BRUNSWICK - UNIT 1 8 3/4 1-3 Amendment No. 127 l
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I tr. ACTIVITY cotrTROL SYSTEM BASES CONTROL ROD PROCRAM CONTROLS (Continued)
The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods wiLL not be withdrawn or inserted.
l The analysis of the rod drop accident is presented in Section 15.4.6 of l
the Updated FSAR and the techniques of the analysis are presented in a topical report (Reference 1) and two supplements (References 2 and 3).
The 18M is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.
3/A.1.5 STAND 8Y LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for maintaining the reactor subcritical in the event that insufficient rods are inserted in the core when a scram is called for. The volume and weight percent of poison material in solution is based on being able to bring the reactor to the suberitical condition as the plant cools to ambient condition. The temperature requirement is necessary to keep the sodium pentaborate in solution. Checking the volume and temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
With redundant pumps and a highly reliable control red scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boren or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
1.
C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BWRs, "C. E. Topical Report NEDO-10527, March 1972.
2.
C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement 1 to NE30-10527, July 1972.
3.
J. A. Haus, C. J. Paone, and R. C. Stirn, addendum 2, "Esposed Cores",
supplement 2 to NEDO-10527, January 1973.
4.
NEDE-24011-P-A, "Ceneral Electric Standard Application for Reactor Fuel,"
Ruvision 6, Amendment 12.
BRUNSWICK - UNIT 1 5 3/4 1-4 Amendment No. 127
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NUCLEAR REGULATORY COMMISSION 4
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CAROLINA POWER & LIGHT COMPANY. et al.
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DOCKET NO. 50-324 i
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.157 License No. DPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendnent filed by Carolina Power & Light Company (thelicensee),datedAugust3,1987,complieswiththestandardsand 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)thattheactivitiesauthorized by this amendment can be conducted witilout endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Cossnission'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 Comission'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-62 is hereby amended to read as follows:
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_2 (2) Technical Specifications The Technical Specifications contained in Ap'pendices A and B, as revised through Amendment No.157, are hereby incorporated in the
' license. Carolina Power & Light Company shall operate the facilit,v in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor Adensam/for Edward A. Reeves, Acting Director Project Directorate II-1
' Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 19,1989 I
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OFFICIAL RECORD COPY
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ATTACHMENT TO LICENSE AMENDMENT NO.157 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 i
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages 3/4 1-14 3/4 1 B 3/4 1-3 8 3/4 1-3 B 3/4 1-4 8 3/4 1-4 l
REACTIVITY CONTROL SYSTEMS 3/4 1.4 CONTROL ROD PROGRAM CONTROLS 100 WORTN MINIMIZER LIMITINC CONDITICA FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be OPERABLE when THERMAL POWER is less than 20% of RATED THERMAL POWER.
APPLICABILITY: Ol8EtATIONAL CONDITIONS 1 and 2*.
l ACTION:
With the RWM inoperable, the provisions of Specification 3.0.4 are not applicable, operation may continue, and control rod movement is permitted provided that a second licensed operator or other qualified member of the technical staff is present at the reactor control console and verifies compliance with the prescribed control rod pattern.
SURVEILLANCE REQUIREMENTS 4.1.4.1.1 The RWM shall be demonstrated OPERA 8LE in OPERATIONAL CONDITION 2, l
prior to withdrawal of control rods for the purpose of making the reactor critical and in OPERATIONAL CONDITION 1 when the RWM is initiated during l
control rod insertion when reducing TWERMAL POWER by:
1.'
Verifying proper annunciation of the selection error of at least one out-of-sequence control rod, and b.
Verifying the rod block function of the RWM by moving an out-of-sequence control rod.
4.1.4.1.2 The RWM shall be demonstrated OPERABLE by verifying the control rod Banked Position Withdrawal Sequence input to the RWM computer is correct l
following any loading of the sequence program into the computer.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is l
permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
B 8RUNSWICK - UNIT 2 3/4 1-14 Amendment No. 157
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REACTIVITY CONTROL SYSTEM i
BASRd CONTROL RODS (Continued) on a scram chan has been analyzed even though control rods with inoperable
(
accumulators may still be inserted with normal drive water pressure.
l Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactors.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in ths FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after each refueling. The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be L.; owed and therefore that other parameters are within their limits, the contcol rod position indication system'must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject' a drive housing.
The required surveillance intervale are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL ROD PROCRAM CONTROLS Cont rol rod withdrawal and insertion sequences are established to assutt that she maximum in sequence individual control rod or control rod segments whi:h are withdrawn at any time during the fuel cycle could noc be worth encugh to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a cancrol rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater than or equal to 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design race of the velocity limiter, could result in a peak enthalpy of 280 cal /gs. Thus, requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than 20%
of RATED THERMAL POWER provides adequate control.
BRUNSWICK - UNIT 2 8 3/4 1-3 Amendment No.157
REACTIVITY CONTROL SYSTEM BASE 3 CONTROL RCD PROGRAM CONTROLS (Continued)
Use of the Banked Position Withdrawal Sequence (BPWS) ensures that in the event of a control rod drop accident, the peak fuel enthalpy will not be greater than 280 cal /gm (Reference 4).
The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.6 of the Updated FSAR, and the techniques of the analysis are presented in a topical report (Reference 1) and two supplements (References 2 and 3).
The EBM is designed to automatically prevent. fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence untd by the operator for withdrawal of control rods.
3/4.1.5 STANDgY LIQUID CONTROL SYSTEM l
1 1
The standby liquid control system provides a backup capability for i
maintaining the reactor suberitical in the event that insufficient rods are i
inserted in the core when a scram is called for. The volume and weight i
percent' of poison material in solution is based on being able to bring the l
reactor to the suberitical condition as the plant cools to ambient condition. The temperature requirement is necessary to keep the sodium pentaborate in solution. Checking the volume and temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
With redundant pumps and a highly reliable control rod scras system, operation of the reactor is permitted to continue for short periods of time l
with the system inoperable or for longer periods of time with one of the redundant components inoperable.
1 1.
C. J. Paone, R. C. Stirn, and J. A. Woodley, " Rod Drop Accident Analysis for Large BNEs " G. E. Topical Report NEDO-10527, March 1972.
2.
C. J. Paone, R. C. Stirn, and R. M. Yound, Supplement i to NEDO-10527, July 1972.
3.
J. 'A. Haus, C. J. Paone, and R. C. Stin, addendum 2 " Exposed Cores" supplement 2 to NEDO-10527, January 197'..
4.
NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel,"
Revision 6, Amendment 12.
l 2RUNSWICK - UNIT 2 8 3/4 1-4 Amendment No.157 i
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