ML20216E605
| ML20216E605 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/07/1999 |
| From: | Campbell G CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20216E600 | List: |
| References | |
| NUDOCS 9909150171 | |
| Download: ML20216E605 (10) | |
Text
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I?ocket Number 50-346 i
License Number NPF-3 Serial Number 2586 Page1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit j
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 2586) concern:
Appendix A, Technical Specifications:
3/4.3.2.1 Safety F:atures Actuation System Instrumentation Bases 3/4.3.1 Reactor Protection System and Safety System Instrumentation and 3/4.3.2 I, Guy G. Campbell, state that (1) I am Vice President - Nuclear of the FirstEnergy Nuclear Operating Company, (2) I am duly authorized to execute and file this certification on behalf of the Toledo Edison Company and The Cleveland Electric Illuminating Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.
By:
o Guy G. C pnpbelT, ViMesident\\ Nuclear Affirmed and subscribed before me this 7th day of September,1999, b
Notary Public, State of Ohio - Nora L. Flood My commission expires September 4, 2002.
9909150171 990907 PDR ADOCK 05000346 P
PDR J
Docket Number 50-346 :
. License Number NPF.
Serial Number 2586 Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specification (TS) 3/4.3.2.1, Safety Features
^ Actuation System Instrumentation, and TS Bases 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety System Instrumentation:
A.
Time Required to Implement: The License Amendment associated with this license amendment application is to be implemented within 120 days after NRC issuance.
B.
Reason for Change (License Amendment Request Number 98-0005):
The proposed changes to TS 3/4.3.2.1 would revise Table 3.3-4, Safety Features Actuation System Instrumentation Trip Setpoints, to delete the " Trip Setpoint" values for Instrument String Functional Unit "b", Containment Pressure - High, and Functional Unit "c", Containment Pressure - High-High, and also modify the
" Allowable Values" entry for these same Functional Units, consistent with updated calculations using current setpoint methodology. The proposed changes would also revise Limiting Condition for Operation (LCO) 3.3.2.1, and Bases 3/4.3.1 and 3/4.3.2 to reflect the removal of the " Trip Setpoint" values for these Functional Units.
C.
Safety Assessment and Significant Hazards Consideration: See Attachment.
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1 Docket Number 50-346 License Number NPF-3 Serial Number 2586 Attachment j
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR 1
LICENSE AMENDMENT REQUEST NUMBER 98-0005 i
(18 pages follow)
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1 LAR 98-0005 Pcge1
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SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 98-0005 TITLE:
Proposed Modification to the Davis-Besse Nuclear Power Station Unit Number 1 j
(DBNPS), Facility Operating License NPF-3, Appendix A - Technical Specifications, to j
Revise Technical Specification (TS) 3/4.3.2.1, Safety Features Actuation System 1
Instrumentation, and the Associated Bases.
DESCRIPTION:
I The proposed TS changes would make certain SFAS instrumentation setpoints consistent j
with updated calculations using current setpoint methodology. The following specific changes are proposed:
Table 3.3-4. Safety Features Actuation System Instrumentation Trio Setnoints The " Trip Setpoint" values for Instrument String Functional Unit "b",
Containment Pressure - High, and Functional Unit "c", Containment Pressure -
High-High, are proposed to be removed from TS Table 3.3-4. Consistent with NUREG-1430, " Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," Revision 1, only the Allowable Value would be specified in the TS for each of these Functional Units. The nominal trip setpoints are specified in the setpoint analysis and are listed in the " Instrument Index," a DBNPS-controlled document, for reference. These two trip setpoints are also listed in the DBNPS Updated Safety Analysis Report (USAR) Section 6.2.4, " Containment Vessel Isolation Systems," Tables 6.2-25, 6.2-26, and 6.2-27.
Future changes to these trip setpoints will be under the regulatory controls of 10 CFR 50.59, " Changes, Tests, and Experiments." These changes will be submitted to the NRC in accordance with the USAR revision requirements of 10 CFR 50.71(c) and the safety evaluation summary report requirements of 10 CFR 50.59(b).
The " Allowable Value" for Instrument String Functional Unit "b", Containment Pressure - High, is proposed to be changed from "s 18.52 psia" to "s 19.38 psia".
The " Allowable Value" for Instrument String Functional Unit "c", Containment Pressure - High-High, is proposed to be changed from "s 38.52 psia" to "s 41.65 psia".
LAR 98-0005 Page 2 In addition, it is proposed to apply the current footnote "##" to the Allowable Values for Functional Units "b" and "c". This footnote signifien St the Allowable Values apply to the Channel Functional Test. The s J: cat footnote,
"#", which signifies that the Allowable Values apply to both the Channel Functional Test and the Channel Calibration, would no longer apply to Functional Units "b" and "c".
Limitinn Condition for Oneration (LCO) 3.3.2.1 The LCO is proposed to be revised to include SFAS Table 3.3-4 Instrunent String Functional Units "b" and "c" in the listing of Functional Units for which only the Allowable Value is specified in the TS.
Bases 3/4.3.1 and 3/4.3.2. Reactor Protection System and Safety System Instrumentation The Bases is proposed to be revised to include SFAS Table 3.3-4 Instrument String Functional Units "b" and "c" in the discussion ofinstrumentation for which only the Allowable Value is specified in the TS.
The proposed changes are shown on the attached marked-up Operating License pages.
SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED:
The Safety Features Actuation System (SFAS) instrumentation trip setpoints and allowable values used to determine the operability of the " Containment Pressure - High" channel and the " Containment Pressure - High-High" channel are affected.
FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:
As described in Section 7.3," Safety Features Actuation System (SFAS)" of the DBNPS Updated Safety Analysis Report (USAR), the function of the SFAS is to automatically prevent or limit fission product and energy release from the core, to isolate the containment vessel, and to initiate operation of the Engineered Safety Features (ESP) equipment in the event of a Loss-Of-Coolant Accident (LOCA). The SFAS consists of four redundant sensing channels and two redundant actuation channels.
Containment Vessel (CV) pressure is one of the station variables monitored by SFAS.
There are four CV pressure trancmitters, one for each SFAS channel. The SFAS also monitors Reactor Coolant (RC) pressure and Borated Water Storage Tank (BWST) level.
L.AR 98 0005 Page 3
'Ihe equipment actuated by SFAS depends upon the severity of the accident, as indicated by the sensor channels. The actuated equipment is separated into five Incident Levels. A brief description of the five SFAS Incident Levels follows. Note that the listings of the actuated equipment are a summary. A complete listing is provided in USAR Figures 7.3-3 through 7.3-8,"SFAS Actuated Equipment Tabulation."
An Incident Level 1 actuation will occur when the RC Pressure - Low or Containment Pressure - High setpoints are reached. The containment purge and containment sampling systems are isolated, the emergency ventilation system is actuated, and the control room ventilation system is isolated.
An Incident Level 2 actuation also occurs when the RC Pressure - Low or i
Containment Pressure - High setpoints are reached. High pressure injection is initiated, the component cooling water system, service water system, containment air coolers, and emergency diesel generators are started, various containment isolation valves are closed, and the containment spray valves are opened (although containment spray pumps are not started).
I An Incident Level 3 actuation occurs when the RC Pressure - Low-Low or Containment Pressure - High setpoints are reached. Low pressure injection is initiated and additional containment isolation valves are closed.
An Incident Level 4 actuation occurs when the Containment Pressure - High-High setpoint is reached. The containment spray system is started and additional containment isolation valves are closed.
An Incident Level 5 actuation occurs when the BWST level setpoint is reached, indicating that the BWST has been nearly depleted. A permissive is generated to allow a manual transfer to the containment emergency sump.
EFFECTS ON SAFETY:
Table 3.3-4. Safety Features Actuation System Instrumentation Trin Setooints Containment Pressure - Hinh The LOCA analysis originally performed to license the DBNPS used a Containment Pressure - High trip setpoint of 4 psig. Summing 4 psig with 14.4 psi as the elevation-corrected conversion factor between gauge and atmospheric pressure yields the current TS trip setpoint of 18.4 psia. It is noted that this value was not instrumentation error corrected.
i Section II.E.4.2, Position 5, of NUREG-0737," Clarification of TMI Action Plan
- Requirements," November 1980, required that the containment pressure setpoint
' Page 4 that initiates containment isolation for nonessential penetrations be reduced to the minimum compatible with normal operating conditions. The DBNPS submitted I
information related to this NUREG-0737 position on January 30,1981 (Toledo Edison letter Serial Number 685). The January,1981 letter stated that the highest expected normal containment pressure is 16.04 psia, and including a 1 psi margin, this results in a setpoint'of 17.04 psia. The letter further stated that, taking into account instrument inaccuracy, the minimum setpoint would be 18.24 psia, and taking into account drift, the minimum Allowable Value would be 18.39 psia. The letter concluded that since these values were very close to the TS trip setpoint and Allowable Values of 18.4 psia and 18.52 psia, respectively, no change was considered necessary. The NRC issued a Safety Evaluation Report closing this issue on April 14,1982 (Toledo Edison Log Number 961).
The Containment Pressure - High setpoint analysis was recently updated in accordance with ISA-S67.04, "Setpoints for Nuclear Safety Related Instrumentation," September 1994, and ISA-RP67.04, Part II, " Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation,"
September 1994, using 17.04 psia as a maximum normal operating limit. As mentioned above, this value is based on the highest expected normal containment i
pressure plus 1 psi margin, considering the guidance of NUREG-0737. Considering instrument string uncertainty, including drift, a trip setpoint of I8.7 psia and a TS Allowable Value for Channel Functional Testing of 19.38 psia were calculated.
An Analytical Limit of 20.83 psia was established based on the new trip setpoint.
The containment response was re-evaluated based on this Analytical Limit, and it was determined that the increase from 18.4 psia to 20.83 psia had a negligible effect on the containment pressure and temperature response following a LOCA or main steam line break (MSLB).
Containment Pressure-Hinh-Hinh The Containment Pressure - High-High trip setpoint is constrained in both
' directions. The trip is set high enough so as to not be activated on a MSLB, such that containment spray is not actuated. The trip is set low enough so that the containment vessel design pressure is not exceeded for the design basis accident.
The current trip setpoint,38.4 psia, is based on a MSLB containment pressure of 24 psig, including margin, plus 14.4 psi to convert to atmospheric. It is noted that this value is not instrumentation error corrected. USAR Section 6.2.1.3.2,
" Containment Pressure Transient Analysis Break Spectrum," gives the peak containment pressure due to a MSLB as 21.4 psig.
i The Containment Pressure - High-High setpoint analysis was also recently updated in accordance with ISA-S67.04, and ISA-RP67.04, Part 11. With the need to include instrument uncertainties, an increased Analytical Limit of 42.8 psia was established.
This value is based on a MSLB containment pressure of 21.4 psig, plus 4 psi margin
LAR 98-0005 Page5 I
I to avoid a containment spray activation, plus 14.4 psi to convert to atmospheric, plus up to 3 psi instrument error correction. The containment response was re-
. evaluated based on the increased Analytical Limit. It was determined that the increase from 38.4 psia to 42.8 psia had a negligible effect on the containment pressure and temperature response following a LOCA or MSLB.
Considering instrument string uncertainty, including drift, a trip setpoint of 40.0 psia and a TS Allowable Value for Channel Functional Testing of 41.65 psia were calculated.
The proposed changes to the Containment Pressure - High and Containment Pressure -
High-High Allowable Values will make the Technical Specifications consistent with i
updated calculations using current setpoint methodology, and have been shown to have only a negligible effect on the containment pressure and temperature response following a LOCA or MSLB. Therefore these changes will have no adverse effect on nuclear safety.
It is noted that although the current TS LCO 3.7.5.1.b limits the ultimate heat sink (UHS) average water temperature to 85 *F, the revised containment performance analysis assumed an initial UHS temperature of 90 'F. This assumption is consistent with the license amendment application submitted on July 28,1999 (FirstEnergy letter Serial Number 2397), currently under NRC review, which proposes to increase the UHS average watec temperature limit to 90 F.
The application of the proposed Allowable Values to only the Channel Functional Test and not the Channel Calibration is consistent with the methodology of NUREG-1430,
" Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," Revision 1, wherein the sensor is calibrated separately from the rest of the instrument string. The proposed removal of trip setpoint values for the containment pressure instrument strings is also consistent with NUREG-1430. Nominal trip setpoints i
are specified in the setpoint analysis and are listed in the " Instrument Index," a DBNPS-controlled document, for reference. These trip setpoints are also listed in the USAR and subject to evaluation under the regulatory requirements of 10 CFR 50.59 prior to changing their values in the future. These are administrative changes and will have no adverse effects on nuclear safety.
Limitine Condition for Ooeration (LCO) 3.3.2.1 The proposed LCO change is associated with the changes to TS Table 3.3-4, and is an administrative change that will have no adverse effect on nuclear safety.
Bases 3/4.3.1 and 3/4.3.2. Reactor Protection System and Safety System Instrumentation The proposed Bases change is associated with the changes to TS Table 3.3-4, and is an administrative change that will have no adverse effect on nuclear safety.
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LAR 98-0005 Pcge 6 i
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. The Davis-Besse Nuclear Power Station (DBNPS) 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 would:
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Not involve a significant increase in the probability of an accident previously evaluated because the proposed changes do not change any accident initiator, i
initiating condition, or assumption.
The proposed changes would revise Technical Specification (TS) Table 3.3-4, Safety Features Actuation System Instrumentation Trip Setpoints, to administratively remove from TS the " Trip Setpoint" values for Instrument String Functional Unit "b", Containment Pressure - High, and Functional Unit "c",
Containment Pressure -Iligh-High, and also modify the TS " Allowable Values" entry for these same Functional Units, consistent with updated calculations using current setpoint methodology. The Trip Setpoint values removed from TS will be maintained in DBNPS-controlled documents. The proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1 and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
Ib.
Not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes do not invalidate assumptions used in evaluating the radiological consequences of an accident, do not alter the source term or containment isolation, and do not provide a new radiation release path or alter radiological consequences.
2.
Not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes do not introduce a new or different accident initiator or introduce a new or different equipment failure mode or mechanism.
3.
Not involve a significant reduction in a margin of safety because the proposed changes establish an error analysis that has been shown to adequately preserve the margin of safety.
LAR 98-0005 Page 7 CONCLUSION:
On the basis of the above, the Davis-Besse Nuclear Power Station has determined that the License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Tecimical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.
ATTACHMENTS:
' Attached are the proposed marked-up changes to the Operating License.
REFERENCES:
1.
DBNPS Operating License NPF-3, Appendix A Technical Specifications through Amendment 233.
2.
DBNPS Updated Safety Analysis Report through Revision 21.
3.
NUREG 1430," Improved Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," Revision 1, April 1995.
4.
DBNPS Calculation C-ICE-48.01-001,"SFAS Containment Pressure Setpoints,"
Revision 6.
5.
DBNPS Calculation C-NSA-060.05-008," Containment Post LOCA Response with Variable SW Temperature," Revision 0, 6.
Instrument Society of America Standards:
ISA-S67.04,"Setpoints for Nuclear Safety Related Instrumentation,"
September 1994.
l ISA-RP67.04, Part 11," Methodologies for the Determination of Setpoints for Nuclear Safety-Related instrumentation," September 1994.
7.
NUREG-0737," Clarification of TMI Action Plan Requirements,"
November 1980.
8.
DBNPS letter to NRC dated January 30,1981 (Toledo Edison Serial Number 685).
9.
NRC letter to DBNPS dated April 14,1982 (Toledo Edison Log Number 961).
10.
DBNPS letter to NRC dated July 28,1999 (FirstEnergy Serial Number 2397).