ML20043G535
| ML20043G535 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/12/1990 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20043G533 | List: |
| References | |
| NUDOCS 9006200403 | |
| Download: ML20043G535 (11) | |
Text
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Proposed Changed Pages i
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Unit 1 Revision i
Page 3/4 1-19 Replace Page 3/4 1-20 Replace Page 3/4 1-22 Replace Unit 2 Revision i
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9006200403 900612 PDR ADOCK 05000348 P
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4 REACTIVITY CORTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERA?;0!1 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position (225 to 231 steps, inclusive)* shall l
be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry vith:
T,,, greater than or equal to 541*F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Vith the drop time of any full length rod determined to exceed the a.
above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
+
b.
With the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricied to less than or equsi to 66% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full longth rods shall be demonstrated through measurement prior to reactor criticality:
a.
For all rods following each removal of the reactor vessel head, b.
For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.
At least once per 18 months.
- The fully withdrawn position used for determining rod drop time shall be greater than or equal to the fully withdrawn position used during subse-quent plant operation.
FARLEY-UNIT 1 3/4 1-19 AMENDMENT NO.
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E REACTIVITY CONTROL SYSTEMS SHUTDOVN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdovn rods shall be fully vithdrawn (225 to 231 steps, inclusive).
APPLICABILITY: MODES 1* and 2*W.
ACTION:
Vith a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour eithert a.
Fully withdrav the rod, or b.
Declare the rod to be inoperable and apply Specification 3.1.3.1.
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SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn (225 to 231 steis, inclusive):
a.
Vithin 15 minutes prior to withdrawal of any rods in control banks 4, B, C or D during an approach to reactor criticality, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
- See Special Test Exceptions 3.10.2 and 3.10.3.
- Vith K,,, greater than or equal to 1.0 e
o FARLEY-UNIT 1 3/4 1-20 AMENDMENT NO.
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FRACTION OF RATED THERMAL POWER Figure 3.11 Red Group insertion Limits Versus Thermal Power Three Loop Operation FARLEY-UNIT 1 3/4 1-22 AMENDMENT NO.
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REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION l
3.1.3.4 The in.iividual full length (shutdovn and control) rod drop time from the fully withdrawn position (225 to 231 steps, inclusive)* shall l
be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
T,,, greater than or equal to $41'F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY:
H0 DES 1 and 2.
ACTION:
a.
With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
b.
Vith the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided TilERMAL POWER is restricted to less than or equal to 66% of RATED TitERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:
a.
For all rods following each removal of the reactor vessel head, b.
For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.
At least once per 18 months.
- The fully withdrawn position used for determining rod drop time shall be-greater than or equal to the fully withdravn position used during subse-quent plant operation.
I FARLEY-UNIT 2 3/4 1-19 AMENDMENT NO.
y REACTIVITY CONTROL SYSTEMS SHUTD0VN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn (225 to 231 steps, inclusive).
APPLICABILITY: MODES 1* and 2*#.
ACTION:
Vith a maximum of one shutdown rod not fully withdrawn, except for t
surveillance testing pursuant to specification 4.1.3.1.2, within one hour either:
a.
Fully withdraw the rod, or b.
Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdovn rod shall be determined to be fully withdrawn (225 to 231 steps, inclusive):
Within 15 minutes prior to withdraval of any rods in control banks a.
A, B, C or D during an approach to reactor criticality, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
- See Special Test Exceptions 3.10.2 and 3.10.3.
- Vith K,,, greater than or equal to 1.0 i
b FARLEY-UNIT 2 3/4 1-20 AMENDHENT NO.
(FULLY WITHDRAWN-225 TO 231 STEPS, INCLUSlVE) 231
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FRACTION OF RATED THERMAL POWER Figure 3.11 Rod Group insertion Limits Versus Thermal Power Three Loop Operction l
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ATTACHMENT 2 i
Significant Hazards Evaluation Pursuant to 10 CFR 50.92 l
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Significant Hazards Evaluation Pursuant to 10 CFR 50.92 for the Proposed Changes to FNP Units 1 and 2 Reactivity Control Systems Technical Specification Proposed Chanae Revise Specifications 3.1.3.4 and 3/4.1.3.5, and Figure 3.1-1 to define the fully withdrawn position for both control and shutdown Rod Cluster Control Assembly (RCCA) banks as between 225 and 231 steps, inclusive.
Backaround The current fully withdrawn position for all of the farley Units 1 and 2 RCCAs, which includes the control and shutdown banks, is 228 steps above rod bottom with a tip to tip distance of 128 steps maintained between the control banks during overlap operation.
" Parking" the RCCAs at the 228 step position for several cycles of operation may cause wear to the RCCA rodlet cladding. This wear is the result of core flow induced vibration which causes contact between the RCCA rodlets and the RCCA guide cards which are located in the upper reactor internals. To avoid chronic wear at the same location on the RCCA rodlet cladding, Westinghouse has recommended to Alabama Power Company for Farley Units 1 and 2 that the fully withdrawn parked position be changed periodically.
By doing so, wear will be spread over a greater surface area of the RCCA rodlet cladding, thus minimizing the probability of complete wear through the cladding at any one spot.
Accordingly, it is proposed to define " fully withdrawn" to mean between 225 and 231 steps, inclusive, above reactor bottom for all RCCA banks.
The tip-to-tip distance between control banks will continue to be maintained at 128 steps during overlap operation.
Between 228 and 231 steps, the RCCAs are withdrawn at least two steps above the active fuel.
Therefore, with respect to core physics, the effects are equivalent.
Additionally, at 231 steps the RCCAs will remain inserted in the guide thimbles of the fuel assemblies and thus will not pose a challenge to the operation of the Control Rod Drive Mechanisms (CRDMs) and also will continue to allow for a smooth rod drop. The assumed control rod drop time considered in the safety analyses will continue to be bounded by thic change.
However, when the RCCAs are withdrawn to 225 steps, they are actually inserted one step (0.63 inches) into the core active fuel.
Consequently, the core physics key safety parameters were evaluated to determine if the change invalidated any safety analysis assumptions.
The effect on the calculstion performed to verify shutdown margin is minimal, resulting in a decrease of 0.03% delta-rho. This effect can be accommodated by the available excess margin at end-of-life which is i
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Page 2 approximately 1.40% delta-rho. Core axial power distributions, differential and integral rod worth as well as other core physics parameters, are also slightly affected; however, sufficient margin exists in the safety analysis to account for these changes. There will also be minimal effect on allowable peaking factors.
Fg is expected to increase by less than 1.0% in the bottom of the core and the axial offset will be l
more negative by less than 1.0%.
Farley has sufficient peaking factor margin to bound both of these effects.
As part of the reload safety evaluation process, the fully withdrawn RCCA position which is selected for use throughout each operating cycle will be evaluated.
Due to the conservatisms incorporated into the Farley safety analyses, the effect of RCCA repositioning is expected to remain bounded for all future reloads.
Analysis Alabama Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed reactor control systems technical specification change and considers this change not to involve a significant hazards consideration, in support of this conclusion, the following analysis is provided:
(1)
The proposed change will not increase the probability or consequences of an accident previously evaluated because RCCA reposition *g will not result in any design or regulatory limits being excmed with respect to the safety analyses documented in the Farley Final Safety Analysis Report (FSAR).
The assumed control rod drop time and reload safety analysis parameters remain bounding.
In addition, since the change does not impact any conditions that would initiate a transient, the probability of previously analyzed events is not increased.
Also, RCCA repositioning will reduce the possibility of rod cladding failure, thereby minimizing the chance of absorber material being introduced into the reactor coolant system.
Expected life of the RCCAs will also be extended.
(2)
The proposed change will not create the possibilit.v cf a new or different kind of accident from any accident pieviously evaluated because the RCCAs will continue to meet their functional requirements.
The RCCAs will remain inserted in the guide thimbles of the fuel assemblies during operation with the proposed withdrawal limits; therefore, the performance of the CRDMs is unaffected by this change. The effect of periodically repositioning the RCCAs is bounded by the Farley safety analysis and therefore will not propagate to a new or different accident.
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I Page 3 (3)
The proposed change will not involve a significant reduction in a margin of safety because RCCA repositioning has an insignificant effect on control rod drop time.
Therefore, rod drop time will continue to be bounded by that assumed in the Farley safety analysis.
There will also be minimal impact on core physics characteristics.
Key safety parameters considered in reactor teloads such as those used in the calculation performed to verify shutdown margin and peaking factors were evaluated and shown not to violate any safety analysis assum)tions.
The effect is negligible
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since the top region of the core :1as such low worth, consequently, the resultant power distribution perturbations are minimal and can be accommodated with available margin. Also, by keeping the tip-to-tip distance during overlap operation the same for the L
proposed definition of fully withdrawn, axial power behavior as a l
function of power level will be maintained.
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Conclusion L
L Based upon the analysis provided herein, Alabama Power Company has determined that the proposed change to the technical specifications will not increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated or involve a reduction in a margin of safety. Therefore, Alabama Power Company has determined that the proposed change meets the requirements of 10 CFR 50.92(c) and does not involve a significant hazards consideration.
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