ML17157A720
| ML17157A720 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/06/1991 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17157A721 | List: |
| References | |
| NUDOCS 9106210129 | |
| Download: ML17157A720 (16) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA PO>IER 5 LIGHT COMPANY ALLEGHENY ELECTP.IC COOPEPATIVE INC.
'2.
DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 109 License No.
NPF-14 The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 5
Light Company, dated january ]8, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's 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 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 set forth.
in 10 CFR Chapter I; 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.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) o'f the Facility Operating License No. NPF-14 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
1o9 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.
PPLL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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This license amendment
',s effective 30 days after its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes o the Technical Specifications Date of Issuance:
June 6, 1991 Walter R. Butler, Director Proiect Directorate I-2 Division of Reactor Projects - I/II
ATTACHMENT TO LICENSE AMENDMENT NO. 109 FACILITY OPERATING LICENSE HO.
NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf pages are provided to maintain document completeness."
REIIOVE 3/4 1-15*
3/4 1-16 B 3/4 1-3 B 3/4 1-4*
INSERT 3/4 1-15*
3/4 1-16 B 3/4 1-3 B 3/4 1-4*
REACTIVITY CONTROL SYSTEHS CONTROL ROD DRIVE HOUSING SUPPORT LIHITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place.
APPLICABILITY:
OPERATIONAL CONDITIONS 1,. 2 and 3.
ACTION:
Nth the control rod drive housing support not in place, be in at least HOT SHUTDON within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE UIREHENTS 4.1.3.8 The control rod drive housing support shall be verified to be in place by a visual inspection prior to.startup, any time it has been disassembled or when aaintenance has been performed in the control rod drive housing support area.
SUSQUEHANNA - UNIT 1 3/4 1-15
REACTIVITY CONTROL SYSTEMS 3/4. 1. 4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM)-shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2", when THERMAL POWER is less than or equal to 20X of RATED THERMAL POWER, the minimum allowable low power setpoint.
ACTION:
a 0 With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed opera-tor or other technically qualified member of the unit technical staff who is present at the reactor control console.
Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.
SURVEILLANCE RE UIREMENTS 4.1.4. 1 The RWM shall be demonstrated OPERABLE:
a.
b.
C.
In OPERATIONAL CONDITION 2 prior to withdrawal of control rods for the purpose of making the reactor critical:
1.
By verifying proper indication of the selection error of at least one out-of-sequence control rod, and 2.
By verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
In OPERATIONAL CONDITION 1 when reducing THERMAL POWER within one hour after reaching the low power setpoint:
1.
By verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod, and 2.
By verifying proper indication of the selection error of at least one out-of-sequence control rod.
By verifying the control rod patterns and sequence input to the RWM computer is correctly loaded following any loading of the program into the computer.
Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is 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.
SUSQUEHANNA - UNIT 1 3/4 1-16 Amendment No. 1.Q9
REACTIVI7Y CONTROL SYSTEMS BASES CONTROL RODS (Continued)
Control rod coupling integrity 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 completing CORE ALTERATIONS that could have affected the control rod coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a con-trol 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 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 ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are 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 homo-
- geneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater than 20K of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm.
Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20X of RATED THERMAL POWER provides adequate control.
The RSCS and RWM logic automatically initiates at the low power setpoint (20K of RATED THERMAL POWER) to provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (which envelope the operating ranges of these variables),
the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably lower than the 280 cal/gm limit.
For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction, and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to deter-mine the peak fuel rod enthalpy rise.
This value is then compared against the SUSQUEHANNA - UNIT 1 B 3/4 1"3 Amendment No. 109
REACTIVITY CONTROL SYSTEMS BASES 3/4. l. 4 CONTROL ROD PROGRAM CONTROLS (Continued) 280 cal/gm design limit to demonstrate compliance for each operating cycle.
If cycle-specific values of the above parameters are outside the range assumed in the parametric
- analyses, an extension of the analysis or a cycle-specific anal-ysis may be required.
Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for perform-ing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume 1.
The RBM 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.
3/4. 1.5 STANDBY LI UID CONTROL SYSTEM The standby liquid control system provides a'backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted.
To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.
A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement.
There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing.
The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41. 2 gpm.
The minimum storage volume of the solution is estab-lished to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.
The temperature requirement for the sodium penetrate solution is necessary to ensure that the sodium penetaborate remains in solution.
With redundant pumps and explosive injection valves and with a highly reliable control rod 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 con-centration will not vary unless more boron 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.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
SUSQUEHANNA - UNIT 1 B 3/4 1-4 Amendment No.
Auo 3 0 1988
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POWER
& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 78 License No. NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 5
Light Company,. dated January 18, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's 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 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 set forth in 10 CFR Chapter I; 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 Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) o'f the Facility Operating License No. NPF-22 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
78 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.
PPKL shall oper ate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective 30 days after its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 6, 1991 Malter R. Butler, Director Project Directorate I-2 Division of Reactor Projects - I/II
C
~
ATTACHMENT TO LICENSE AtlEtlDVEt'l0. 78 FACIL'ITY OPERATItln LICENSE l!0. tlPF-22 DOCKET VO. 50-388 Replace the following paces of the Appendix A Tcchnical Specifications with enclosed pages.
The revised pages are identif',ed by Amendment number and contain vertical lines irdicating the area. of change.
The overleaf pages are provided to mairtain document completeness.*
REROVE 3/4 1-15*
3/4 1-16 B 3/4 1-3*
B 3/4 1-4 INSERT 3/4 1-15*
3/4 1-16 B 3/4 1-3*
B 3/4 1-4
.REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE HOUSING SUPPORT
. LIMITING CONDITION FOR OPERATION 3..1.3.8 The control rod drive housing support shall be in place.
APPLICABILITY:
OPERATIONAL CONDITIONS'1, 2 and 3.
ACTION:
With the control rod drive housing support not in place, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE UIREMENTS
- 4. 1.3.8 The control rod drive housing support shall be verified to be in place by a visual inspection prior to startup any time it has been disassembled or when maintenance has been performed in the control rod drive housing support area.
SUSQUEHANNA - UNIT 2 3/4 1-15
REACTIVITY CONTROL SYSTEMS 3/4. 1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2", when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER, the minimum allowable low power setpoint.
ACTION:
a 0 With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console.
Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.
SURVEILLANCE RE UIREMENTS
- 4. 1.4. 1 a.
b.
C.
The RWM shall be demonstrated OPERABLE:
In OPERATIONAL CONDITION 2 prior to withdrawal of control rods for.
the purpose of making the reactor critical:
1.
By verifying proper indication of the selection error of at least one out-of-sequence control rod, and 2.
By verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
In OPERATIONAL CONDITION 1 when reducing THERMAL POWER within one hour after reaching the low power setpoint:
1.
By verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod, and 2.
By verifying proper indication of the selection error of at least one out-of-sequence control rod.
By verifying the control rod patterns and sequence input to the RWM computer is correctly loaded following any loading of the program into the computer.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is 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.
SUSQUEHANNA - UNIT 2 Amendment No.
78
REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) reactor scram and will isolate the reactor coolant system from the containment when required.
Control rods with inoperable accumulators are declared inoperable and Spe-cification 3. 1.3.
1 then applies.
This prevents a pattern of inoperable accumu-lators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be in-serted 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 reactor.
Control rod coupling integrity 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 completing CORE ALTERATIONS that could have affected the control rod coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
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 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 aormal 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 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 ROD PROGRAM CONTROLS
'ontrol rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are 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 20X of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm.
Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER provides adequate control.
SUSQUEHANNA " UNIT 2 B 3/4 1"3,
Amendmen No. 31
~&up
REACTIVITY CONTROL SYSTEMS BASES CONTROL ROD PROGRAM CONTROLS (Continued)
The RSCS and RWM logic automatically initiates at the low power setpoint (20K of RATED THERMAL POWER) to provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (which envelope the operating ranges of these variables),
the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably lower than the 280 cal/gm limit.
For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction, and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to de-termine the peak fuel rod enthalpy rise.
This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle.
If cycle-specific values of the above parameters are outside the range assumed in the parametric
- analyses, an extension of the analysis or a cycle-specific analysis may be required.
Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume 1.
The RBM 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.
3/4. 1.5 STANDBY LI UID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free
- shutdown, assuming that none of the withdrawn control rods can be inserted.
To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.
A minimum quantity of 4587 gallons of sodium pentaborate solution containing.a minimum
'f 5500 lbs. of sodium pentaborate is required to meet this shutdown require-ment.
There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing.
The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm.
The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.
The temperature requirement for the sodium penetrate solution is necessary to ensure that the sodium penetaborate remains in solution.
With redundant pumps and explosive injection valves and with a highly reliable control rod 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.
SUS(UEHANNA - UNIT 2 B 3/4 1-4 Amendment No. 78