ML20113G849
| ML20113G849 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/30/1992 |
| From: | Myers T, Shelton D CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20113G846 | List: |
| References | |
| 2022, NUDOCS 9205130002 | |
| Download: ML20113G849 (7) | |
Text
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Docket Number 50-346 i,
Licens.e Number NPF-3 Serial Number 2022 Enclosure Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 Attached are requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1, Facility Operating License Number NPF-3.
Also included is the Safety Assessment and Significant Hazards Consideration.
The proposed changes submitted under cover letter Serial Number 2022 concern:
Appendix A, Technical Specification 5.3.2, Reactor Core - Control Rods Fort D. C. Shelton, Vice President, Nuclear - Davis-Besse p
By:
T. ),
ers Ditector - Technical Services Sworn and Subscribed before me this 30th day of Apri?., 1992.
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- 10 0 Notary Pub /i'c, State of Ohio EVELYNL DRESS "W PUDUC,STE OF WO E@tsJuly 28,1994 920S130002 920430 PDR ADOCK 05000346 P
Docket Number-50-346 License Number NPF-3 Serial Number 2022 Enclosure Page 2 The following inforr.ation is provided to support issuance of the requested changes to the Davis.Besse Nuclear Pover Station, Unit Number 1, Operating Licence Number NPF-3, Appendix A Technical Specifications (TS), TS 5.3.2, Reactor Core - Control Rods.
A.
Time Required to Implement This change is to be implemented prior to startup irom the eighth refueling outege.
B.
Reason for change (License Amendment Request 91-0025): This license amendment request proposes revision of TS 5.3.2 to allow the use of extended life control roda and the use of axial power shaping roos of different Inconel absotber material.
C.
Safety Assessment and Significant Hazards Consideration:
See attached.
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Dockat Number-50-346.
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License Number NPF-3:
Serial Number 2022 Attachment Page 1-SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATI01 FOR LICENSE AMENDHENT REr"EST 91-0025 TITLE:
Revision to Technical Specification (TS) 5.3.2, Reactor Core - Control Rods, to Provide a Description of Extended Life Control Rods and of Axial Power Shaping Rods.
DESCRIPTION:
This License Amendment Request proposes a revision of TS 5.3.2 to includa a descriptionuof extended life control rods and to allow flexibility in the description of the axial power shaping rods (APSRs).
The description of the safety and regulating control rods that is currently given in TS 5.3.2 applies to the B&V Fuel Company standard Mark-B control rods.- There are small differences between the new extended isfe control rods and the Mark-B control rods.
The silver-indium-cadmium neutron absorber in the new extended life control rods has a slightly smaller diameter, which is offset by the absorber being 5 inches longer. Also, the extended life control rods have Inconel cladding, in place of the stainless steel cladding (type 304) used for the Mark-B design.
Davis-Besse Nuclear Power Station, Unit 1 (DBNPS) plans to use eight extended life control rod assemblies (LLCRAs) in operating Cycle 9 and additional ELCRAs in following cycles. The ELCRAs are designed for a 22 effective full power year lifetime. The request to use ELCRAs is similar to that ap;Toved by the
' Nuclear Regulatory Commission (NRC) for the Florida Power Corporation's Crystal River 3 Nuclear Generating Plant as Amendment 103 issued on Dececher 14,.1987 (TAC No. 65259).
This License Amendment Request also proposes a revision to the description'of AFSRs in TS 5.3.2 by changing the description of the absorber material from inconel-600 to lnconel. Although at the present time DBNPS has no plans to replace the APSRs, this change vill allow future design. flexibility.
SYSTEMS, COMPONENTS, AND ACTIVITIES-AFFECTED:
This TS revision affects.only the control rods by allowing the use of extended life control rods and by allowing design flexibility.in the alloy used.for the absorber in the APSRs.
SAFETY FUNCTIONS OF-THE AFFECTED SYSTEMS. COMPONENTS, AND ACTIVITIES:
The-safety function of the-control rods is.to shut down the reactor.
In addition,.the. control rods control fast reactivity changes, and, in conjunction with the APSRs, burnable poison rods and soluble boron, control lifetime reactivity and power distribution changes.
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License Number NPF-3 i
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Page 2 EFFECTS ON SAFETY:
The ELCRAs are the same design as those previously approved by the NRC.
They have been used at other BW designed 177 fuel assembly facilities.
No problems have been observed with the ELCRAs.
As described above, the ELCRAs are not significantly dii'ferent from the Mark-B control rod assemblies that are currently in use at the DBNPS.
The neutron absorber in the ELCRAs has a slightly smaller diameter than that for the Mark-B design (0.386 inch versus 0.392 inch). This slight decrease in diameter is offset by a longer absorber (139 inches versus 134 inches). The resulting worth for the ELCRA design is equal to that of-the standard Mark-B-design at the beginning of the cycle and is slightly greater at the end of cycle. The ELCRAs vill normally be
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loaded symmetrically in the core to pre *:ent any small induced quadrant power tilt due to the slightly different reactivity vorth charc.cteristics of:the ELCRAs compared to the standard Mark-B design.
The' rod worths for a particular fuel cycle are evtluated as part of the reload, analyses to ensure that they are acceptable for the unmntrolled control rod assembly group vithdrawal accidents (for a vithdrawal from a suberitical condition and for a withdrawal at power), for th_e control
-rod assembly misalignment accident (stuck-out, stuck-in, or dropped control rod assembly), for the control rod assembly ejection accident, and for all other accidents which involve reaci.or trips / shutdown. The requirements on the rod worths also ensure that there is no reduction l
in a margin of safety.
l The extended life control rods have a larger absorber-to-cladding gap, thicker cladding, and different cladding material (Inconel versus stainless steel) in order to increase the lifetime. The increased absorber-tc-cladding Fap allows for additional avelling of the absc,rber that results from the allowed increase in the thermal neutron fluence.
Calculations for the extended life control rods show that design l
considerations are met for clad creep collapse, stren, fatigue, and stroking _vear.
Inconel has very low corrosion rates compared to type
'304 stainless steel,-and the effect of corrosion-on mechanical margins is considered insignificant for the extended life control-rods.
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Although the activatior of the ELCRA cladding and the subrequent corrosion and wear of the cladding vill be different from that of the Mark-S design, there vill be a negligible change in the overall radioactive Fource term in the reactor coolant from all sources.
For interchangeability, all ELCRA interface dimensions with the control rod drive mechanism, reactor internals, and handling equipment are kept the same as.the Mark-B design.
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_The external dimensions of the~ELCRAs are effectively ioen*ical to the l
Mark-B design.
The overall length of the extended life con:rol rods is the same as that for the-Mark-B design.
Also, the external diameter of the cladding (with tolerance included) for the extended life control rods is the same as that for the Mark-B design. The extended life control rods are designed with adequate flexibility, as well as clearances, to permit freedom of motion within the fuel assembly guide 1
Docket Number 50-346 License Number NPF-3
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-Page 3 tubes throughout the stroke. The ELCRAs veigh the same as the Mark-B control' rod assenblies.
Because of these ELCRA characteristics, control rod drop times should be unaffected, and rod insertion rate requirements are expected to be met.
Control rod drop time surveillance testing, required prior to startup, vill verify this.
Thermal hydraulic analyses were performed on the ELCRA design.
Due to the design characteristics of the ELCRA, the maximum absorber temperatures for the ELCRA during normal operation, anticipated operational occurrences, and accidents-are significantly less than the melting temperature of the absorber. The internal pressure of the ELCRA is maintained below nominal system pressure during normal operation and anticipated' operational occurrences. This provides conservative assurance of control rod insertability and subsequent withdrawal. The. design characteristics of the ELCRA ensure that the ELCRAs are insertable and the cladding integrity is maintained during normal cperation, anticipated operational occurrences, and accidents.
Calculations show that bulk boiling vill not occur in the guide tube ll annulus, even with very conservative assumptions, such as crediting a l
very low guide tube flovrate compared to the actual predicted flovrate.
L Consequently, the thermal-hydraulic analysis for the extended life control rod design is acceptable.
The only proposed change in the description of the APSRs in TS 5.3.2 is the change of the absorber material from Inconel-600 to Inconel. This absorber is 63 inches long and is encased and cealed in stainless steel cladding.- Normal design controls vill preclude the use of types of
'Inconel that are not appropriate. The determination of the acceptability of another type of Inconel vould include a calculation of the irradiation swelling and thermal expansion of the absorber to be rure that the cladding strain from the radial expansion is acceptable and that there is sufficient axial space for the growth. Also, the internal pressure, absorber temperature, and. cladding stress vould be determined as part of the normal ~ design process to show that they are acceptable. The acceptability, in terms of physical design and neutronics performance, of-any change in the absorber material vould be addressed in the vendor design control process'and in the safety evaluation for the reload report that corresponds to the first use of differcnt abrorber material. Accordingly, the use of a different absorber material in the APSRs vill be controlled so that there vill
-not be an increase in the potential for fuel damage or isotopic releases.
Based on the above, it is concluded that approval by the NRC of this License Amendment Request vill have no adverse effect on safety.
-SIGNIFICANT HAZARDS CONSIDERATION:
The NRC has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility.
A propcsed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would:
(1) Not involve a significant increase in the probability or consequences of an accident t
Docket Numbar 50-346.
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prevluusly evaluated; - (2) Not create the possibility of a new or different kind of accident from any previously evaluated; or (3) Not involve a significant reduction in a margin of safety. Toledo Edison has-reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS) in accordance with the proposed changes vould la.
Not= involve a significant increase in the probability of an accident previously evaluated, because the physical design parameters for the extended life control rod assemblies (ELCRAs),
as described previously, assure that there is no increase in the probability of a stuck control rod assembly. The changes to the ELCRAs do not affect the initiation of any other accident previously evaluated in the DBNPS Updated Safety Analysis Report.
Also, any change in the absorber material for the axial power shaping rods (APSRs) vill be analyzed prior to use to ensure that the physical design parameters are acceptable from a safety perspective.. Therefore, the effects of these changes will not affect the initiation of a previously evaluated accident and, therefore, vill not lead to a significant increase in the probability of an accident.
Ib
- Not involve a significant increase in tha consequences of an accident previously evaluated, because there are no changes in the design that could result in more serious fue) damage or isotopic releases than previously evaluated. The rod worths for each fuel cycle are evaluated as part of the reload analyses to ensure that they are within the bounds of previously evaluated accidents, as
-discussed above.
Also, the mechanical and thermal-hydraulic analyses for the ELCRAs show that-acceptable performance of the control rods will be maintained. Thus, the radiological
-consequences of an accident previously evaluated cre not increased by the change to the ELCRAs. In addition,=any change in the absorber material for the APSRs vould be analyzed for acceptability before use to ensure that the radiological consequences are not increased by the change.
~2a.
Not create the possibility of a new kind of accident from any accident.previously_ evaluated, because the differences between the Mark-B design and-the ELCRA design have been evaluated to ensure that all changes are acceptable as described above. The changes are minor and do not create any new accident initiators.
- Also, any change in the absorber material for the APSRs vould be analyzed prior to use for acceptability as described above.
Therefore, the effects of these changes will not lead to the initiation of a new kind of accident.
2b.
Not create the possibility of a different kind of accident from any. accident previously evaluated, because the ditferences between the Mark-B design and the ELCRA design have been evaluated to ensure that all changes are acceptable as described above. The changes are minor and do not create any new accident initiators.
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Docket Nuabst 50-346 License Number NPP-3 Serial Number 2022 Attachment Page 5 Also, any change in the absorber material for the APSRs vould be analyzed prior to use for acceptability as described above.
Therefore, the effects of these changes vill not lead to the initiation of a different kind of accident.
3.
Not involve a significant reduction in a margin of safety since the changes from the Mark-B design to the ELCRA design are
-relatively minor, and the effects of the changes have all been evaluated, as discussed above, to ensure that safety margins are not significantly affected.
Also, the effects of any future changes in the absorber for the APSRs vould be evaluated prior to use to be sure that safety margins are not significantly affected.
CONCLUSION:
On the basis of the above, Toledo Edison has determined that the License Amendment Request does not involve a significant hazards consideration. As:this License Amendment Request' concerns proposed changes to the Technical Specifications that must be reviewed by the NRC, this License Amendment Request does not constitute an unreviewed safety question.
ATTACHMENT Attached are the propoced' marked-up changes to the Operating License.
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