ML20078Q098
| ML20078Q098 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/06/1994 |
| From: | Stetz J CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20078Q102 | List: |
| References | |
| 2251, GL-93-05, GL-93-5, NUDOCS 9412210156 | |
| Download: ML20078Q098 (14) | |
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300 Madison Avenue John P. Stetr Vice President - Nuclear Davis-Besse Docket Number 50-346 License Number NPF-3 Serial Fumber 2251 December 6, 1994 United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C. 20555 Subj ect : Proposed Modifications to the Davis-Besse Nuclear Power Station (DBNPS) Operating License Technical Specifications Pursuant to Generic Letter 93-05, "Line-Item Technical Speci-fication Improvements to Reduce Surveillance Requirements for Testing During Power Operation" Gentlemen:
Enclosed is an application for an amendment to the DBNPS Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications (TS). This application proposes the modification of TS 3/4.1.3.1,
" Group Height - Safety and Regulating Rod Groups;" TS 3/4.4.6.2, "Oper-ational Leakage;" TS 3/4.5.2, "ECCS Subsystems - Tavg equal to or greater than 280*F;" TS 3/4.6.2.1, " Containment Spray System;" and TS 3/4.10.4, " Shutdown Margin" in accordance with the recommendations of Gene.:ic Letter (GL) 93-05. In addition, a change is proposed to TS 3/4.1.3.2, " Group Height - Axial Power Shaping Rod Group."
The proposed change to TS 3/4.1.3.2 is a related change not addressed by GL 93-05. The proposed deletion of Surveillance Requirement 4.1.3.2.2 (that presently requires the movement at least 2% of each Axial Power Shaping Rod not fully withdrawn) every 31 days has been removed from the requirements of NUREG-1430, " Improved Standard Technical Specifications for B&V Plants." Therefore, this change is also submitted as a line item improvement.
All proposed Technical Specification changes are compatible with DBNPS operating experience and are consistent with NRC guidance provided by GL 93-05 or NUREG-1430. Toledo Edison requests that these changes be approved by the NRC by May 31, 1995.
\4CCCB Operating Companies Cleveland Electne muminating Toledo Edison 9412210156 941206 -
lI PDR ADOCK 05000346 - I P PDR
j Docket Number 50-346 License Number NPF-3 Serial Number 2251 Page 2 Should you have any questions or require additional information, please contact Mr. Villiam T. O'Connor, Manager - Regulatory Affairs, at (419) 249-2366.
Very truly yours, FVK/laj Enclosure cc L. L. Gundrum, DB-1 NRC/NRR Project Manager J. B. Martin, Regional Administrator,.NRC Region III S. Stasek, DB-1 NRC Senior Resident Inspector J. R. Villiams, Chief of Staff, Ohio Emergency Management l Agency, State of Ohio (NRC Liaison) l Utility Radiological Safety Board 1
Dockat Nu ber 50-346 License Number NPF-3
, Serial Number 2251 Enclosure Page 1 j APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 l 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 l Consideration.
The proposed changes (submitted under cover letter Serial Number 2251) concern:
Appendix A, Technical Specifications 3/4.1.3.1. " Group-lielght - Safety -
and Regulating Rod Groups;" 3/4.1.3.2, " Group Height - Axial Power Shaping Rod Groups" 3/4.4.6.2, " Operational Leakage " 3/4.5.2, "ECCS Subsystems - Tavg equal to or greater than 280'F;" 3/4.6.2.1,
" Containment Spray Systems" and 3/4.10.4, " Shutdown Margin."
By:
J. P. ftetz', Vi6e Pr'esident - Nuclear
, #[
Sworn and subscribed before me this 6th day of December, 1994.
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Dockat Nu:ber 50-346 l L License Number NPF-3 i Serial Number 2251 '
Enclosure Page 2 The following information is provided to support issuance of the j requested changes to Davis-Besse Nuclear Power Station, Unit Number 1 ;
Operating License Number NPF-3, Appendix A, Technical Specifications. ;
^ The changes involve Technical Specifications 3/4.1.3.1, " Group Height - !
Safety and Regulating Rod Groups;" 3/4.1.3.2, " Group Height - Axial ^
Power Shaping Rod Group;" 3/4.4.6.2, " Operational Leakage;" 3/4.5.2, "ECCS Subsystems - Tavg equal to or greater than 280'F;" 3/4.6.2.1,
" Containment Spray System;" and 3/4.10.4, " Shutdown Margin." !
A. Time Required to Implement: This change is to be implemented !
within 90 days after the NRC issuance of the License Amendment. ,
B. Reason for Change (License Amendment Request 94-0007): ;
1 The purpose of the proposed changes is to modify the DBNPS !
Operating License NPF-3 by adopting the appropriate surveillance test reductions recommended by Generic Letter (GL) 93-05, -
"Line-Item Technical Specification Improvements to Reduce !
Surveillance Requirements For Testing During Power Operation." i Technical Specification 3/4.1.3.2, " Group Height - Axial Power ! '
Shaping Rod Group" is also proposed for modification to conform to the requirements of NUREG-1430, " Improved Standard Technical .
Specifications for B&W Plants." !
C. Safety Assessment and Significant Hazards Consideration See l Attachment. ;
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Dockct Number 50-346 i Lic nsa Nurber NPF-3 >
.S:rici Nu ber 2251- -
.' Attachment 1 !
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION f FOR LICENSE AMENDMENT REQUEST 94-0007 (32 pages follow) !
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1 LAR 94-0007 Page 1 l SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION '
FOR LICENSE AMENDHENT REQUEST NO. 94-0007 TITLE:
Proposed Modification to the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications to Revise the following Technical Specifications: 3/4.1.3.1, " Group Height - Safety and Regulating Rod Groups;" 3/4.1.3.2, " Group Height - Axial Power Shaping Rod Group;" 3/4.4.6.2,
" Operational Leakage;" 3/4.5.2, "ECCS Subsystems - Tavg equal to or greater than 280"F;" 3/4.6.2.1, " Containment Spray System;" and 3/4.10.4, " Shutdown Margin."
DESCRIPTION:
The purpose of the proposed changes is to modify the DBNPS Operating License NPF-3 by adopting the appropriate surveillance test reductions recommended by Generic Letter (GL) 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements For Testing During Power Operation." Technical Specification 3/4.1.3.2, " Group fleight - Axial Power Shaping Rod Group" is also proposed for modification to conform to the requirements of NUREG-1430,
" Improved Standard Technical Specifications for B&V Plants."
The Nuclear Regulatory Commission (NRC) Staff completed an examination of surveillance requirements in Technical Specifications (TS) that require testing during power operation. This effort was a part of the NRC Technical Specifications Improvement Program. The results of this work are reported in NUREG-1366, " Improvements to Technical Specifications Surveillance Requirements," December 1992.
The NRC Staff examined TS surveillance requirer- is in order to identify those that should be improved. The NRC Staff evaluai sn of the safety benefit of changes to the TS surveillance requirements involved the consideration of the purpose of the surveillance requirement (how a change affects safety, including the reduction of challenges to plant systems), the effect that the performance of the surveillance requirement has on personnel (the exposure of personnel to radiation sources and the burden on personnel resources), and the effect that the performance of the surveillance requirement has on plant equipment (equipment wear or degradation). In performing this study, the NRC Staff found that, while the majority of the testing at power is important, safety can be improved, equipment degradation decreased, and an unnecessary burden on personnel resources eliminated by reducing the amount of testing that certain TS require during power operation.
All proposed Technical Specification (TS) changes are compatible with DBNPS operating experience and are consistent with the guidance provided by Generic Letter 93-05 or NUREG-1430.
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LAR 94-0007 Page 2 Change Number 1: Control Rod Hovement Test The proposed change to TS 3/4.1.3.1, " Group Height - Safety and Regulating Rod Groups" Surveillance Requirement 4.1.3.1.2, will change the 31 day surveillance test iniorval to 92 days for demonstrating the operability of each safety and regulating control rod which is not fully inserted. Reference Section 4.2.1 of Enclosure 1 to GL 93-05.
The purpose of this change is to extend the surveillance test interval from 31 days to 92 days. This is desirable because performance of this surveillance test by moving the rods at least 2% can potentially cause reactor trips, dropped rods, and unnecessary challenges to safety systems. This change is acceptable because the purpose of the control rod movement test is to detect rods that cannot move, yet industry-experience shows that most stuck rods are discovered in plant startup during initial pulling of the rods or during rod drop testing and not during the performance of this surveillance test.
Change Number 2: APSR Hovement Test The proposed change to TS 3/4.1.3.2, " Group Height - Axial Power Shaping Rod Group" will delete Surveillance Requirement 4.1.3.2.2, which states "Unless all APSR are fully withdrawn, the APSR shall be determined to be OPERABLE by moving the APSR rods at least 2% at least once every 31 days."
Reference:
TS 3.1.6 of NUREG-1430, " Improved Standard Technical Specifications for B&V Plants."
The purpose of this change is to remove the requirement to move the APSRs at least 2% at least once every 31 days. This requirement does not appear in the Improved Standard Technical Specifications. It was removed in Revision 0 of NUREG-1430. This change is acceptable because the APSRs have no safety function to insert in order to mitigate any design bas *s accident. In fact, by the design of the system, these rods are prevented from moving on a reactor trip.
Since these rods reside in an area of high neutron flux during operation, allowing them to move during a trip could have the effect of adding positive reactivity. This proposed change is a line item improvement.
Change Number 3: Control Rod Insertion Test The proposed change to TS 3/4.10.4, " Shutdown Margin," Surveillance Requirement 4.10.4.2 vill increase the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period during which to perform the safety and regulating control rod full insertion test to 7 days. Reference Section 12 of Enclosure 1 to GL 93-05.
The purpose of this change is to allow the use of special test exception TS 3/4.10.4 vithout performing another rod insertion test as long as TS 3/4.10.4 is entered within 7 days of performing Surveillance Requirement 4.1.3.4.c. The control rods are tripped under Surveillance Requirement 4.1.3.4c (Rod Drop Time) prior to startup to verify that control rod drop times are less than the value assumed in the safety analyses. If TS 3/4.10.4 is invoked more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the rod insertion test of Surveillance Requirement 4.1.3.4.c another rod insertion test would need to be performed under the current requirements of Surveillance Requirement 4.10.4.2. Invoking the special test exception of TS 3.4.10.4 vithin the 7 days allowed by the proposed change without performing an additional rod insertion test is acceptable as long as there are no changes to the core or control rods after the first test.
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- LAR 94-0007 Page 3 i Change Number 4: Reactor Coolant System Pressure Isolation Valves l The proposed change to TS 3/4.4.6.2, " Operational Leakage," Surveillance Requirement 4.4.6.2.2b will increase the 72-hour time for remaining in Cold Shutdown (Mode 5) before requiring leak testing of the Reactor Coolant System (RCS) Pressure Isolation Valves (PIVs) CF-30, CF-31, DH-76 and DH-77 to 7 days. '
References Section 6.1 of Enclosure 1 to GL 93-05.
The purpose of this change is to incorporate the recommendations of NUREG-1366 that the 72-hour time for remaining in cold shutdown without testing the RCS P1Vs for leaks be increased to 7 days. This vill allow Toledo Edison more time to perform unrelated repairs during a short shutdown without unduly increasing the restart test scope and occupational exposure, and vill have an insignificant impact on risk.
Change Number 5: Containment Debris Inspections i
The proposed change to TS 3/4.5.2, "ECCS Subsystems - Tavg greater than or equal to 280*F," Surveillance Requirement 4.5.2.c.2, vill revise the text to state, "Each ECCS subsystem shall be demonstrated OPERABLE. . . by a visual inspection. . . For containment entries, during the period when CONTAINMENT INTEGRITY is required, by inspecting, at least once daily, the areas of containment af fected by the entry, and during the final exit (containment closeout)."
Reference:
Section 7.5 of Enclosure 1 to GL 93-05. This proposed change in wording differs slightly from the wording recommended by GL 93-05 in order to provide clearer direction consistent with the terminology utilized at the DBNPS.
The purpose of this change is to implement the ALARA concept. A visual inspection of the containment is made to assure that no loose debris is present which could be transported to and clog the containment emergency sump. This change vill reduce the potential for radiation dose to people performing containment inspections when containment integrity is established. Requiring inspections at the completion of each containment entry results in a higher dose than necessary to personnel performing the inspections. Inspecting the affected areas of the containment at least once daily if the containment has been entered .
that day, and during the final exit is sufficient to ensure that there is no !
loose debris that would clog the containment emergency sump.
Change Number 6: Containment Spray System Nozzles ,
The proposed change to TS 3/4.6.2.1, " Containment Spray System," Surveillance Requirement 4.6.2.1.d vill change the 5 year surveillance interval of the Containment Spray System to 10 years.
Reference:
Section 8.1 of Enclosure 1 to GL 93-05.
The purpose of this change is to allow an extension of the surveillance interval to 10 years. This is acceptable because industry experience indicates that the only problems found at pressurized water reactors under this test have been construction-related problems. Furthermore, this testing gives no qualitative data on flow rates exiting the spray nozzles. It only verifles that there is flow, which from the industry-experience data, is not a problem.
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LAR 94-0007 Page 4 SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED:
The proposed Technical Specification changes affect the testing of:
Safety and Regulating Rod Groups.
Axial Power Shaping Rod Groups.
Reactor Coolant System (RCS) Pressure Isolation Valves (PIVs). These are:
the Core Flooding Tank to Reactor Check Valves CF-30 and CF-31 and the Decay Heat /Lov Head Safety Injection Pump Discharge to Reactor Coolant System Stop Check Valves DH-76 and DH-77.
Containment Spray System nozzles.
For the proposed Technical Specification change affecting the inspection requirements for containment debris, the Decay Heat /Lov Pressure Injection Pumps and the Containment Spray Pumps are indirectly affected. These pumps can be affected because they have the capability of taking suction from the containment emergency sump. Surveillance Requirement 4.5.2.c.2 requires that the containment be inspected after containment entry, when containment integrity is in effect, for debris that may be washed into the emergency sump during an accident.
FUNCTIONS OF THF, AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:
Change Numbers 1 and 3: Control Rod Movement and Insertion Tests The control rods are used during normal operations to control reactor core reactivity during power operations. Their movement in and out of the core in combination with changes of the dissolved boron concentration of the reactor coolant provide for the capability of power level changes in response to electrical system demands.
The rod control system is described in the DBNPS Updated Safety Analysis Report (USAR) Section 7.4.1.1, " Control Rod Drive Control System (CRDCS) - Trip Portion." The safety function directly affected by this change is that portion of the system which alious the rods to fall into the reactor core when the monitored Reactor Coolant System parameters reach their trip setpoints. When the trip setpoints are reached, power is removed from the Control Rod Drive Mechanisms (CRDHs) and the rods fall by gravity into the core. The control rods are inserted into the core upon receipt of the RPS, Anticipatory Reactor Trip System (ARTS), Diverse Scram System (DSS), or manual trip signals, which act to de-energize the CRDMs. The reactor shutdown feature (the trip portion) of the CRPCS is the only aspect of the CRDCS that affects public safety.
Change 1 is a test to ensure rods are free to move while at power without shutting down. Change 3 is to ensure all rods can be inserted to ensure positive shutdown while doing control rod vorth testing with shutdown margin less than 1% ok/k.
LAR 94-0007 i Page 5 Change Number 2: APSR Hovement Test The APSRs are positioned in areas of high neutron flux to reduce excess power production in those areas, thus providing a lower peak power while average power remains at the desired level. The APSRs are described in USAR Section 4.2.3.1.1.2. The function of the testing activity is to demonstrate the capability of the APSRs to move, however, the capability of this movement is not required for a design basis accident.
Change Number 4: Reactor Coolant System Pressure Isolation Valves The safety function of the Reactor Coolant System Pressure Isolation Valves CF-30, CF-31, DH-76, and DH-77 is to prevent an intersystem LOCA. The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the Decay Heat Removal (DHR) System outside of containment.
The accident is the result of a postulated failure of the PIVs, which are part of the reactor coolant pressure boundary, and the subsequent over pressurization of the DHR System downstream of the PIVs from the RCS. Leakage through these valves is included as a part of the RCS identified leskage. This ensures that any leakage that may occur will be limited to an amount that does not compromise safety. Monitoring this leakage serves as an indication of the condition of these pressure isolation valves and provides trend data that can be used to provide information on possible degradation of the valves.
The four pressure isolation valves are used during normal DHR system operation as the return path for DHR cooling. Valves DH-77 and CF-31 are in the return path for DHR Loop 1 and valves DH-76 and CF-30 are in the return path for DHR Loop 2. The function of the testing is to ensure that, after these valves bue ,
opened to provide shutdown cooling, the valves properly reseat and are leak tight.
Change Number 5: Containment Debris Inspections There is no direct safety function associated with performing a containment valkdown to ensure that no loose debris is present. This walkdown indirectly affects safety because the Decay Heat /Lov Pressure Injection and Containment Spray pumps take a suction on the containment emergency sump during the recirculation phase of a Loss Of Coolant Accident (LOCA). The purpose of the walkdown is to ensure there is no loose debris that could be washed into the '
containment emergency sump and block the suction path during the recirculation phase of a LOCA.
Change Number 6: Containment Spray System Nozzles The Containment Spray System is described in the DBNPS Updated Safety Analysis Report (USAR) Section 6.2.2.2.2, Containment Spray System. The containment Spray System is an engineered safety feature which has the dual function of removing heat and fission product iodine from the post-accident containment atmosphere. The system serves no function during normal operation.
Removal of heat is accomplished by directing borated water spray into the containment. The system consists of two pumps, two spray headers, isolation valves, and the necessary piping, instrumentation and controls. The pumps take
LAR 94-0007 Page 6 suction initially from the Borated Vater Storage Tank and later from the '
containment emergency sump during the recirculation phase. The Containment Spray System shares the Borated Vater Storage Tank (BVST) and the suction lines from the tank with the high and low pressure injection systems.
The containment spray nozzles are installed on two containment ring her.ders. ,
Each header has 90 nozzles, for a total of 180 nozzles. .
The function of the testing activity is to ensure that the flow path from the pump discharge through the spray nozzles is not blocked.
EFFECTS ON SAFETY: ,
Change Number 1: Control Rod Movement Test The proposed change to the DBNPS Technical Specification 3/4.1.3.1, " Group Height - Safety and Regulating Rod Groups" will extend the surveillance test interval for Surveillance Requirement 4.1.3.1.2 from 31 days to 92 days. The proposed change is in conformance with the NRC recommendations of Generic Letter 93-05, Enclosure 1, Section 4.2.1. The DBNPS reliability data for the control Rod Drive Control System has been reviewed and supports extending the surveillance frequency in accordance with the conclusions of Generic Letter 93-05. Therefore, there is no adverse effect on safety by changing the surveillance test interval.
Change Number 2: APSR Movement Test The proposed change to the DBNPS Technical Specification 3/4.1.3.2, " Group '
Height - Axial Power Shaping Rods" will remove the requirement to move the APSRs at least 2% at least once every 31 days. This requirement no longer appears in the Standard Technical Specifications. It was removed in Revision 0 of NUREG-1430. This change is acceptable because the APSRs have no safety function to insert in order to mitigate any design basis accident. In fact, by the design of the system, these rods are prevented from moving on a reactor trip.
Since these rods reside in an area of high neutron flux during operation, allowing them to move during a trip could have the effect of adding positive reactivity. The DBNPS reliability data for the movement of APSRs has been reviewed and supports the removal of this Surveillance Requirement from the TS '
which is consistent with the Improved Standard Technical Specifications. The remainder of the existing Technical Specification vill be retained which requires that the APSRs be properly aligned to assure proper power distribution {
vithin design power peaking limits. There is no adverse effects on safety by '
deleting the requirements of this surveillance.
Change Number 3: Control Rod Insertion Test i
The proposed change to the DBNPS Technical Specification 3/4.10.4, " Shutdown l Hargin" vill increase the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period during which to perform the I rod full insertion test to 7 days. The proposed change is in conformance with the NRC recommendations of Generic Letter 93-05. Enclosure 1, Section 12. As discussed above, the DBNPS reliability data for the Control Rod Drive Control System has been reviewed and supports extending the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period during I
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LAR 94-0007 i Page 7 which to perform the rod full insertion test to 7 days in accordance with the conclusions of Generic Letter 93-05. Therefore, this change causes no adverse effect-on safety.
i Change Number 4: Reactor Coolant System Pressure Isolation Valves l The proposed change to the DBNPS Technical Specification 3/4.4.6.2, " Operational Leakage" will extend the surveillance interval for Surveillance Requirement 4.4.6.2.2.b from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. The proposed change is in conformance with the NRC recommendations of Generic Letter 93-05, Enclosure 1, Section 6.1. The DBNPS surveillance data collected as a part of the RCS. identified leakage surveillance testing indicates.that leakage through these valves has not been a recent problem at the DBNPS. Furthermore, Surveillance Requirement 4.4.6.2.1.d !
requires performance of a Reactor Coolant System vater inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation. This inventory balance vill provide ongoing indication of the integrity of these RCS Pressure Isolation Valves. In addition, these valves vill continue to be leak tested each i refueling outage and are included in the DBNPS Check Valve Reliability Program !
which is designed to identify early signs of valve degradation. . Extension of this surveillance frequency in accordance with GL 93-05 vill produce no adverse effect on safety.
Change Number 5: Containment Debris Inspections '
The proposed change to the DBNPS Technical Spec lication 3/4.5.2, "ECCS Subsystems - Tavg greater than 280'F," Surveillance Requirement 4.5.2.c.2 changes the requirement for a containment vs1kdown of affected areas after each entry when containment integrity is established, to once per day and during the final exit for containment closeout when containment integrity is established.
This change does not make substantial changes to-these requirements nor does it !
change the method of performing the surveillance. However, this change vill-limit the potential for radiation dose to people performing containment inspections when containment integrity is established. This change has no adverse effect on safety.
Change Number 6: Containment Spray System Nozzle The proposed change to the DBNPS Technical Specification 3/4.6.2.1,
" Containment Spray System," will extend the surveillance interval for Surveillance Requirement 4.6.2.1.d from 5 years to 10 years. The proposed 'l change is in conformance with the NRC recommendations of Generic Letter 93-05, Enclosure 1, Section 8.1. N'JREG-1366 discusses this surveillance in Section 8.1. It notes that in preparing NUREG-1366, the NRC Staff searched for problems involving the containment spray system that had been uncovered by means of this surveillance testing. Only three cases were found and in all three cases the problem involved a construction trror. At the DBNPS this surveillance test has been satisfactorily performed in the past. Because this system is not used during normal operation there is no mechanism by which these nozzles could be rendered nonfunctional. Therefore, the extension of the surveillance interval from 5 years to 10 years vill cause no adverse effect on safety.
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LAR 94-0007 Page 8 SIGNIFICANT HAZARDS CONSIDERATION:
The Nuclear Regulatory Commission 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 proposed 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 previously evaluated; (2) Not create the possibility of a new or different kind of accident from any accident 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 No. 1 in accordance with these changes vould la. Not involve a significant increase in the probability of an accident previously evaluated because no change is being made to any accident initiator. The proposed changes to extend the surveillance test intervals for control rod exercising and containment spray nozzle testing, to increase the allovable time period for conducting control rod drop testing prior to startup, to remove the requirements to move the APSRs at least 2%
every 31 days, to increase the time allowed in cold shutdown prior to requiring leak testing the Reactor Coolant System Pressure Isolation Valves, and to reduce the number of containment valkdowns to identify debris are the incorporation of the recommendations of Generic Letter 93-05, NUREG-1366, or the Improved Standard Technical Specifications (NUREG-1430). Generic Letter 90-06 and NUREG-1366 contain the results and recommendations of the NRC Staff's study on determining which TS surveillance requirement testing can be safely reduced. This study and its recommendations have been reviewed by Toledo Edison and found applicable to the DBNPS. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.
Ib. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes to extend the surveillance test intervals for control rod exercising and containment spray nozzle testing, to increase the allovable time period for conducting control rod drop testing prior to startup, to remove the requirements to move the APSRs at least 2% every 31 days, to increase the time allowed in cold shutdown prior to requiring leak testing the Reactor Coolant System Pressure Isolation Valves, and to reduce the number of containment l
valkdowns to identify debris do not invalidate accident conditions or I assumptions used in evaluating the radiological consequences of an accident. l i
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- 2. Not create the possibility of a new or different kind of accident from any '
accident previously evaluated because the proposed changes to extend the surveillance test intervals for control rod exercising and containment spray nozzle testing, to increase the allovable time period for conducting control rod drop testing prior to startup, to remove the requirements to move the APSRs at least 2% cvery 31 days, to increase the time allowed in
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LAR 94-0007 I Page.9 cold shutdown prior to requiring leak testing the Reactor Coolant' System Pressure Isolation Valves, and to reduce the number of containment- ,
valkdowns to identify debris do not change the way the plant is operated. l No new. types.of-failures or accident initiators are introduced by the proposed changes. !
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- 3. Not involve a significant reduction in a margin of safety because the !
proposed changes to extend the surveillance test intervals for control rod ;
exercising and containment spray nozzle test ng, to increase the allowable , ;
time period for conducting. control rod drop :esting-prior to startup,1to remove the requirements to move the APSRs at least 2% every 31 days, to increase the time allowed in cold shutdown prior to requiring leak testing the Reactor Coolant System Pressure Isolation Valves, and to reduce the i number of centainment walkdowns to identify debris have no adverse effect I on the' operation of the systems used to mitigate design bases accidents i and therefore, there is no significant reduction in the margin of safety.
I conclusion l
i on the basis of the above, Toledo Edison has determined'that the License !
-Amendment Request does not involve a significant hazards consideration. As this l License Amendment Request concerns a proposed change to the Technical ,
Specifications that must be reviewed by the Nuclear Regulatory Commission, this 3 License Amendment Request does not constitute an unreviewed safety question.
Attachment Attached are the proposed marked-up changes to the Operating License.
References !
NUREG-1366 " Improvements to Technical Specifications Surveillance Requirements," }
December 1992. i 5
NUREG-1430 " Improved Standard Technical Specifications for B&V Plants" i i
Generic Letter 93-05 "Line Item. Technical Specifications Improvements to Reduce i Surveillance Requirements For Testing During Power Operations." '
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Davis-Besse Nuclear Power Station Updated Safety Analysis Report Sections t 4.2.3.1.1.2, 6.2.2.2.2 and 7.4.1.1. ,
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