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Latest revision as of 19:33, 6 October 2021
| ML20217N439 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/27/1998 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217N445 | List: |
| References | |
| RTR-NUREG-1431 2508, NUDOCS 9803090024 | |
| Download: ML20217N439 (12) | |
Text
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9" LW Davig,BeSSe Nuclear Pdwer Station E EE 5S01 North State Route 2 m Oak Harbor Ohio 43449-9760 John K. Wood 4I9u249-2300 Mce President . Nuclear Fu: 419-321-8337 Docket Number 50-346 License Number NPF-3 Serial Number 2508 February 27, 1998 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001
Subject:
Supplemental Information Regarding License Amendment Application to Revise Technical Specifications Regarding the Control Room Emergency Ventilation System Limiting Condition for Operation (LAR 97-0005; TAC No. M98521)
Ladies and Gentlemen:
The purpose of this letter is to provide supplemental information for a pending license amendment application regarding the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1, Operating License Number NPF-3. The subject application, License Amendment Request (LAR) Number 97-0005, dated April 18,1997 (Toledo Edison Serial Number 2447), involves proposed changes to Technical Specification (TS) 3/4.7.6, Plant Systems - Control Room Emergency Ventilation System (CREVS), and the associated Bases 3/4.7.6, Control Room Emergency Ventilation System. /
This supplemental information is provided in response to a verbal request from the NRC Staff to modify the proposed new Action 3.7.6.1.b to make it more consistent with NUREG-1431, " Improved Standard Technical Specifications for Westinghouse Pressurized Water Reactors," Revision 1. Associated with this change, TS Bases 3/4.7.6 /
has also been modified. These changes have been appropriately incorporated 9 to the enclosed Safety Assessment and Significant Hazards Consideration. Revision t trs are included in the right margin to denote the changes.
9903090024 990227 PDR ADOCK 05000346'
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. Docket Number 50-346 License Numbe,r NPF-3 Serial Number 2508
' Page 2 l I
Additional supplemental information was previously provided by letter dated October 10, .
1997 (Toledo Edison Serial Number 2487), in response to a verbal request from the NRC Staff to explain a variance between the proposed changes and NUREG-1430," Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reacters,"
Revision 1. l
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Should you have any questions or require additional information, please contact Mr. James L. Freels, Manager - Regulatory Affairs, at (419) 321-8466.
Ve truly yours, M aj Enclosure cc: A. B. Beach, Regional Administrator, NRC Region III A. G. Hansen, DB-1 NRC/NRR Project Manager {
S. J. Campbell, DB-1 NRC Senior Resident Inspector l J. R. Williams, Chief of Staff, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
Utility Radiological Safety Board l
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. Docket Number 50-346 License,Numbqr NPF-3
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l Serial Number 2508 Enclosure Page1
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i APPLICATION FOR AMENDMENT l TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION l
UNIT NUMBER 1 i
l Attached are the 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.
i The proposed changes (submitted under cover letter Serial Number 2508) concern:
Appendix A, Technical Specification Section 3/4.7.6, Plant Systems - Control Room Emergency Ventilation System, and the associated 'Bases, j
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By: -#
John K. Wo[d, Vice President - Nuclear I Sworn to and subscribed before me this 27th day of February,1998
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d Notary Public, State of Ohio L. AURA A. JEtMSON l Notary Pubbe, State of Ohio My Commission Empires 8-15 2001 l
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1 Docket Number 50-346 License,Numbqr NPF-3 Serial Number 2508 )
Enclosure Page 2 The following information is provided to support issuance of the requested changes to the i Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1, Facility Operating License Number NPF-3, Appendix A, Technical Specifications. The changes involve Technical Specifications (TS) 3/4.7.6, Plant Systems - Control Room Emergency Ventilation System,-
and the associated Bases 3/4.7.6, Control Room Emergency Ventilation System.
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 Requer.t Number 97-0005):
This application proposes to revise the TS Limiting Condition for Operation (LCO) to include new required Actions in the event that one or both channels of Station Vent Normal Range Radiation Monitoring instrumentation become inoperable.
These radiation monitoring instrumentation channels serve to isolate time control room normal ventilation system in the event of high radiation levels in the station vent stack. The associated TS Bases are also proposed to be revised ::onsistent with the proposed changes.
The proposed changes will eliminate the potential for an unnecessary plant shutdown in the event both channels of radiation monitoring instrumentation become inoperable during plant operation.
C. Safety Assessment and Significant Hazards Consideration: See Attachment l
. Docket Number 50-346 License Number NPF-3 Serial N' umber 2' 508 Attachment SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NO. 97-0005 (12 pages follow) i
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- LAR 97-0005, Revision 1 Page1, .
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION .
FOR
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LICENSE AMENDMENT REQUEST NO. 97-0005, REVISION 1 '
TITLE: '
Proposed Modification to the Davis-Besse Nuclear Power Station, Unit Number 1, Facility Operating License NPF-3, Appendix A, Technical Specifications to Revise the Cont.al Room Emergency Ventilation System (CREVS) Limiting Condition for Operation.
DESCRIPTION:
The purpose of the proposed changes is to modify the Davis-Besse Nuclear Power Station
' (DBNPS) Operating License NPF-3, Appendix A, Technical Specifications (TS) and the associated Bases. The proposed changes would revise the TS Limiting Condition for Operation i
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(LCO) to include new required Actions in the event that one or both channels of Station Vent Normal Range Radiation Monitoring instrumentation become inoperable. These radiation monitoring instrumentation channels serve to isolate the control room normal ventilation system in the event of high radiation levels in the station vent stack. The associated TS Bases are also proposed to be revised consistent with the proposed changes.
The proposed changes are described in further detail as follows:
. TS 3/4.741 Plant Systems - Control Room Emereency Ventilation System i
The present Action 3.7.6.1 would be renumbered as Action 3.7.6.1.a, and Actions 3.7.6.1.b and 3.7.6.1.c would be added, to read as follows:
- b. With one channel of Station Vent Normal Range Radiation Monitoring instrumentation inoperable, restore the inoperable channel to OPERABLE status, or isolate the control room normal ventilation system and place at least one control room emergency ventilation system train in operation within 7 days,
- c. With both channels of Station Vent Normal Range Radiation Monitoring l instmmentation inoperable, within I hour, isolate the control room normal ventilation system and place at least one control room emergency ventilation system train in operation.
l
. LAR 97-0005, Revision 1 Page 2 .
These proposed changes will eliminate the potential for an unnecessary plant shutdown in the event both channels of radiation monitoring instrumentation become inoperable during plant operation, by requiring that at least one CREVS train be placed in operation. Currently, should both channels of station vent radiation monitoring instrumentation become inoperable, the plant must enter TS 3.0.3 and commence a shutdown within one hour.
Proposed Action 3.7.6.1.c would also bring the DBNPS TS into closer agreement with NUREG-1430, " Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," Revision 1. Babcock and Wilcox (B&W) Improved Standard Technical Specification (ISTS) LCO 3.3.16, " Control Room Isolation - High Radiation," requires that the CREVS be
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placed in the emergency recirculation mode within I hour in the event the radiation monitoring !
instrumentation channel becomes inoperable in Modes 1 through 4. It is noted that for the B&W I ISTS, the control room isolation signal is provided by a single channel, while the DBNPS design has two channels.
Proposed Action 3.7.6.1.b is modeled after NUREG-1431, " Improved Standard Technical Specifications for Westinghouse Pressurized Water Reactors," Revision 1, as discussed during a conference call with the NRC staff on January 14,1998. Westinghouse ISTS LCO 3.3.7 is based on a two channel design and includes an Action statement which allows 7 days to place a train of control room emergency ventilation in service in the event a single inoperable channel of radiation monitoring instrumentation cannot be restored.
Consistent with the terminology utilized in the new Action statements, Surveillance Requirement (SR) 4.7.6.1.e.2 would be revised to change " Station Vent Radiation High" to " Station Vent Normal Range Radiation Monitoring."
Bases 3/4.7.6 Control Ro6m Emercency Ventilation System 1
Consistent with the proposed changes to TS 3/4.7.6.1, Bases 3/4.7.6 would be modified by adding two additicnal paragraphs, to read as follows:
The Station Vent Normal Range Radiation Monitoring isolation function provides that under the required conditions, an isolation signal will be given. The Station Vent Normal Range Radiation Monitors provide isolation and shutdown of the control room nonnal ventilation system.
Under the proposed Action statements for inoperable Station Vent Normal Range Radiation Monitoring instrumemation, should the control room normal ventilation system be isolated and at least one train of the control room emergency ventilation system be placed in operation, these systems would be in a state equivalent to that which they would be in following an actual high radiation condition. Plant operation can continue indefinitely in this state, provided that control room temperature can be maintained in an acceptable range, with the control room emergency ventilation system obtaining fresh-air makeup as described in the Updated Safety Analysis Report Section 9.4.1, " Control Room."
LAR 97-0005, Revision 1 Page 3 .
1 Each of the proposed changes is shown on the attached marked-up Operating License pages.
SYSTEMS. COMPONENTS. AND ACTIVITIES AFFECTED:
The affected systems, components, and activities are the Control Room Emergency Ventilation System (CREVS), the Control Room Normal Ventilation System, the Station Vent Normal Range Radiation Monitoring instmmentation channels, and the required TS actions to be taken in the event these instmmentation channels are inoperable.
l FUNCTIONS OF THE AFFECTED SYSTEMS. COMPONENTS, AND ACTIVITIES: l The CREVS is described in Section 9.4.1, " Control Room," of the USAR. The CREVS is designed to provide filtered and conditioned air to the control room. This ensures that control room equipment and habitability for control room personnel remain unaffected during and following all credible accident conditions. Operability of the CREVS,in conjunction with other control room design provisions, is based on limiting control room personnel exposure consistent !
with the requirements of General Design Criteria 19, " Control Room," of Appendix A of 10 CFR 50.
Each section of the normal control room ventilation system that penetrates the control room l pressure boundary is equipped with two (redundant) safety-related, pneumatically-operated ;
isolation dampers which function to isolate the ductwork from the control room upon receipt of a i Safety Features Actuation System (SFAS) Incident Level I signal. An SFAS Incident Level I signal is currently initiated as a result of high radiation level within the containment building, low reactor coolant pressure, or high pressure within the containment building. SFAS Actuation Channel 1 isolates one of the redundant trains of dampers, and SFAS Actuation Channel 2 isolates the other train. Each train of damper supply air is equipped with a pressure switch which senses low control air pressure to th~ e dampers and initiates a station computer alarm and stops the corresponding train of control room air handling unit's (supply) fan and the control room return air fan. The CREVS is put into operation manually.
In addition to the SFAS Incident Level I signal, a Station Vent Normal Range Radiation Monitoring isolation signal from radiation monitors RE4598AA or RE4598BA will also automatically isolate the control room and shutdown the corresponding train of normal control room ventilation. These independent radiation monitors rneasure station vent discharge, and upon detection of high radiation, isolate the control room in a manner similar to the SFAS Incident Level 1 signal described above (RE4598AA similar to SFAS Actuation Channel 1 and RE4598BA similar to SFAS Actuation Channel 2). These radiation monitors continuously monitor the station vent during plant operational Modes 1 through 4 for particulate, iodine, and noble gas radioactivity. These monitors are discussed in Section 11.4.2.2.4, " Unit Vent Monitoring System," and Table 11.4-1, " Liquid, Gas, and Airborne Radiation Monitors," of the USAR. The station vent is the final discharge point for the gaseous radioactive effluents.
I Area radiation monitors are provided for the control room which continuously give the background radiation level. In case of any abnormal increase in the background level, the control 1 room operator can manually isolate the normal ventilation system and start the CREVS. i l
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LAR 97-0005, Revision 1 Page 4 .
EFFECTS ON SAFETY:
The proposed change to LCO 3.7.6.1 would provide new required Actions in the event that one or both channels of Station Vent Normal Range Radiation Monitoring instrumentation become inoperable. In the event that one channel is inoperable, the proposed Action statement would require that the control room normal ventilation system be isolated and at least one CREVS train be placed in operation within 7 days. The proposed allowable outage time of 7 days for one inoperable channel allows a reasonable time period for corrective actions to repair the equipment or to isolate the control room normal ventilation system and place at least one train of CREVS in operation should repair efforts be unsuccessful. In addition, this time is consistent with the present allowable outage time for one inoperable CREVS train. The proposed Action to place at least one CREVS train in operation within one hour, in the event both channels of radiation monitoring become inoperable,is more conservative than the present Action which requires that a plant shutdown commence within one hour, but does not require the CREVS be placed in operation. Since completion of these required actions places the ventilation systems in a state equivalent to that which occurs were a high radiation isolation to occur, this proposed change will have no adverse effect on plant safety.
With the control room isolated and CREVS in operation, continued plant operation is contingent upon maintaining control room temperature (as measured in the Cabinet Room)in an acceptable range. In accordance with Plant Procedure DB-OP-02533, " Control Room Emergency Ventilation System Load Shedding," if the temperature in the Cabinet Room reaches 105*F, a rapid plant shutdown is performed, and if the temperature in the Cabinet Room reaches 110 F, the reactor is manually tripped.
The proposed change to the terminology utilized in SR 4.7.6.1.e is an administrative change made to make the terminology consistent with the proposed new Action statements. This change will have no adverse effect on plant safety.
The proposed changes to Bases 3/4.7.6 are administrative changes consistent with the proposed ;
changes to LCO 3.7.6.1, and will have no adverse effect on plant safety.
SIGNIFICANT H AZARDS 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. The Davis-Besse Nuclear Power Station has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No.1, in l
i accordance with this change would:
=
LAR 97-0005, Revision 1 Page 5 .
la. Not involve a significant increase in the probability of an accident previously evaluated because no accideut initiators, conditions, or assumptions are affected by the proposed changes.
The proposed change to Limiting Condition for Operation (LCO) 3.7.6.1 would include new required Action statements in the event that one or both channels of Station Vent Normal Range Radiation Monitoring instrumentation become inoperable. Under the proposed Action statements for inoperable Station Vent Normal Range Radiation Monitoring instrumentation, should the control room normal ventilation system be isolated and at least one train of the control room emergency ventilation system be placed in operation, these systems would be in a state equivalent to that which they would be in following an actual high radiation condition. These proposed changes have no bearing on the probability of an accident.
The proposed change to the terminology utilized in Surveillance Requirement (SR) 4.7.6.1.e is an administrative change made to make the terminology consistent with the proposed new Action statements. The proposed changes to Bases 3/4.7.6 are administrative changes consistent with the proposed changes to LCO 3.7.6.1. These changes have no bearing on the probability of an accident.
Ib. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes do not change the source term, containment isolation, or allowable releases.
As described above, under the proposed Action statements for inoperable Station Vent Normal Range Radiation Monitoring instmmentation, should the control room normal ventilation system be isolated and at least one train of the control room emergency ventilation system be placed in operation, these systems would be in a state equivalent to that which they would be in following an actual high radiation condition. Therefore, in the unlikely event of an accident requiring control room isolation while in this condition, the dose consequences to control room operators would be unchanged.
The proposed change to the terminology utilized in Surveillance Requirement (SR) 4.7.6.1.e is an administrative change made to make the terminology consistent with the proposed new Action statements. The proposed changes to Bases 3/4.7.6 are administrative changes consistent with the proposed changes to LCO 3.7.6.1. These changes have no bearing on the consequences of an accident.
- 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed changes.
As described above, under ie proposed Action statements for inoperable Station Vent Normal Range Radiation Monitoring instrumentation, should the control room normal ventilation system be isolated and at least one train of the control room emergency ventilation system be placed in operation, these systems would be in a state equivalent to I
LAR 97-0005, Revision 1 Page 6 that which they would be in following an actual high radiation condition. Operation of the l equipment and components in this manner would not introduce the possibility of any new or different kinds of accidents.
The proposed change to the terminology utilized in Surveillance Requirement (SR) l 4.7.6.1.e is an administrative change made to make the terminology consistent with the {
proposed new Action statements. The proposed changes to Bases 3/4.7.6 are administrative changes consistent with the proposed changes to LCO 3.7.6.1. These changes would not introduce the possibility of any new or different kinds of accidents.
- 3. Not involve a significant reduction in a margin of safety because the proposed changes to the Action under LCO 3.7.6.1 ensure that control room isolation capability is maintained in the event a station vent radiation monitor is inoperable. The proposed allowable j' outage time of seven days for one inoperable channel is consistent with the presently allowable outage time for one inoperable CREVS. The proposed Action to place at least one CREVS train in operation within one hour, in the event both channels of radiation monitoring become inoperable, is more conservative thu the present Action which requires that a plant shutdown commence within one hour, but does not require the ,
CREVS be placed in operation.
The proposed change to the terminology utilized in Surveillance Requirement (SR) 4.7.6.1.e is an administrative change made to make the terminology consistent with the proposed new Action statements. The proposed changes to Bases 3/4.7.6 are administrative changes consistent with the proposed changes to LCO 3.7.6.1. These changes would not affect the margin of safety.
CONCLUSION:
1 On the basis of the above, the Davis-Besse Nuclear Power Station has determined that the l License Amendment Request does not involve a significant hazards consideration.
ATTACHMENT: l l
Attached are the proposed marked-up changes to the Operating License.
REFERENCES:
through Revision 20.
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LAR'97-0005, Revision 1 Page 7 . .
1
- 5. NUREG-1430, Revision 1, " Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors."
- 6. NUREG-1431, Revision 1, " Improved Standard Technical Specifications, for Westinghouse Pressurized Water Reactors."
- 7. 10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants," )
" Criterion 19 - Control Room."
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