ML20080G266

Application for Amend to License NPF-3,providing New RCS pressure-temp Limits Curves Applicable Up to 21 EFPYs

ML20080G266
Person / Time
Site:	Davis Besse
Issue date:	01/30/1995
From:	Stetz J CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20080G269	List: ML20080G266 ML20080G275
References
2269, NUDOCS 9502070079
Download: ML20080G266 (11)

Text

- GNTENf0R  !

W ENERGY 300 Madison Avenue # P 8" Toledo, OH 43652-0001 Vice President - NucWir 419-249-2300 Dovis Besse i

Docket Number 50-346 License Number NPF-3 Serial Number 2269 l

January 30, 1995 ,

United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C. 20555 Subj ect: Proposed Modification to the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications to Revise Technical Specifications 3/4.4.9.1, Pressure-Temperature Limits Reactor Coolant System, Figures 3.4-2, 3.4-3, and 3.4-4, Bases Section 3/4.4.9, and License Condition 2.C(3)(d).

Gentlemen:

Enclosed is an application for an amendment to the DBNPS Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications. The e proposed changes involve Technical Specification (TS) 3/4.4.9.1 Pressure-Temperature Limits, Reactor Coolant System, Figures 3.4-2, 3.4-3, and 3.4-4, Bases Section 3/4.4.9, and License Condition 2.C(3)(d).

This application proposes changes to the above documents to provide new Reactor Coolant System pressure-temperature limit curves that are applicable up to 21 effective full power years (EFPY).

The current curves are applicable up to 10 EFPY vhich is projected to be reached on September 1, 1995. Therefore, to allow for training and implementation, Toledo Edison requests that the NRC approve and issue these changes by August 1, 1995.

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i Should you have any questions or require additional information, please  !

contact Mr. William T. O'Connor, Manager - Regulatory Affairs, at l (419) 249-2366. i i

Very truly yours, j FVK/laj l i

ect L. L. Gundrum, DB-1 NRC/NRR Project Manager J. B. Martin, Regional Administrator, NRC Region III l S. Stasek, DB-1 NRC Senior Resident Inspector .

J. R. Villiams, Chief of Staff, Ohio Emergency Management  !

Agency, State of Ohio (NRC Liaison)  !

Utility Radiological Safety Board i

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Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NPF-3 DAVIS-BESSE NUCLEAR POWER 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 2269) concern:

Appendix A, Technical Specification 3/4.4.9.1, Pressure-Temperature Limits, ,

Reactor Coolant System, Figures 3.4-2, 3.4-3, and 3.4-4, Bases Section 3/4.4.9, and License Condition 2.C(3)(d).  ;

By: _

J. P. SWETY,' Vic'e'PVesident - Nuclear Sworn and subscribed before me this 30th day of January.1995.

Notary Public, State of Ohio EVELYN L DRESS Notary Public, Siste of Ohio My Commission Empires 7/28/99 I

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.* Liccnr2 Nurber NPF-3 Seriel Nu:bar 2269 Enclosure Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications. The changes involve Technical Specifications 3/4.4.9.1 Pressure-Temperature Limits, Reactor Coolant System, Figures 3.4-2, 3.4-3, and 3.4-4, Bases Section 3/4.4.9, and License Condition 2.C(3)(d).

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 Number 94-0006):

This application proposes changes to the TS that provides new Reactor Coolant System pressure-temperature limit curves that are applicable up to 21 effective full power years (EFPY). The present Technical Specification pressure-temperature limit curves are applicable up to 10 EFPY.

C. Safety Assessment and Significant Hazards Consideration: See Attachment

.- l SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 94-0006 TITLE A proposed change to the Davis-Besse Nuclear Power Station, Unit 1 Operating License Appendix A, Technical Specification 3/4.4.9.1, Pressure-Temperature Limits Reactor Coolant System, Figures 3.4-2, 3.4-3 and 3.4-4, Bases Section 3/4.4.9 and License Condition 2.C(3)(d).

DE,SCRIPTION The proposed Technical Specification provides new Reactor Coolant System pres-sure-temperature limit curves which are applicable up to 21 effective full power feers (EFPY). The present Technical Specification pressure-temperature limit curves are applicable up to 10 EFPY. The following Technical Specification figures are revised by this change:

Figure 3.4-2 Reactor Coolant System Pressure-Temperature Limits for Heatup and Criticality for the First 10 EFPY Figure 3.4-3 Reactor Coolant System Pressure-Temperature Limits for Cooldown for the First 10 EFPY Figure 3.4-4 Reactor Coolant System Pressure-Temperature Heatup and Cooldown Limits for Inservice Leak and Hydrostatic Tests for the First 10 EFPY Technical Specification Bases Section 3/4.9 is being revised to reflect the extension to 21 EFPY of the pressure-temperature limits.

This proposed Technical Specification change also extends the applicability of the Technical Specification figures 3.4-2a and 3.4-2b from 10 EFPY to 21 EFPY.

License condition 2.C(3)(d) is being revised to reflect the applicability of analyses to ensure protection against lov temperature overpressure events to 21 EFPY, SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED Reactor Coolant System / Reactor Vessel Pressure-Temperature Limits FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS AND ACTIVIT7;ES The reactor vessel is an integral part of the reactor coalant pressure boundary (RCPB). It is essential that the material integrity of the reactor vessel be maintained. The reactor vessel prersure-temperature limits are provided in the l Technical Specifications to ensure operation of the plant is in compliance with j these limits.

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EFFECTS ON SAFETY

Background

The pressure-temperature limits of the RCPB are established in accordance with the requirements of Appendix G to 10 CFR 50 " Fracture Toughness Requirements."  !

The limitations prevent non-ductile failure during normal operation, including l anticipated operational occurrences and system hydrostatic tests. The loading l conditions of interest include: ,

1. normal operations, including heatup and cooldown, l

2. inservice leak and hydrostatic tests, and

3. reactor core operation. l The major components of the RCPB have been analyzed in accordance with Appen-dix G to 10 CFR 50. The closure head region, reactor vessel outlet nozzles and the beltline region have been identified to be the only regions of the reactor l vessel, and consequently of the RCPB, that determine the pressure-temperature limitations concerning non-ductile failure.

To demonstrate compliance with the fracture toughness requirements, Appendix G to 10 CFR 50 requires that the reactor vessel ferritic materials be tested in accordance the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and that the reactor vessel beltline materials be tested in accordance with the requirements of 10 CFR 50 Appendix H, " Reactor Vessel Material Surveillance Program Requirements." The effects of neutron radiation on '

the reference nil-ductility temperature (RT of the reactor vessel beltline materials, inE.) and the upper shelf energy (USE)luding the veld ,

predicted from the results of this surveillance program. The surveillance  !

program at Davis-Besse complies with the requirements of Appendix H to 10 CFR 50 l and is described in the B&W Owners Group Topical Report BAV-1543A " Master Integrated Reactor Vessel Material Surveillance Program." The surveillance program at Davis-Besse consists of six surveillance capsules. Each of the capsules contains test specimens fabricated from two base metals, a veld metal, two heat-affected-zone metals and a correlation metal. Surveillance capsules TEl-F, TEl-B, TEl-A and TEl-D vere removed at the end of Cycles 1, 3, 4 and 6, respectively. The results of the analysis of the surveillance results are given in BAV-1701, BAV-1834, BAV-1882 and BAV-2125.

The RT is an index by which the reactor vessel material fracture toughness properTh,isassessedinaccordancewiththerequirementsofASMECodeSection III, Appendix G. The effects of neutron induced embrittlement on the reactor vessel beltline materials are determined by trending the increase in RT as a function of the applied neutron fluence. The changes in the RT are E11tored by testing of the surveillance samples as part of the Reactor Vessel Surveil-lance Program. The increase in the RT is added to the initial RT to produce an adjusted RT,g which serves" N input into the fracture meEEisnics analysis.

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• / Liesns2 Numb 2r NPF-3 S:rici Numbar 2269 Attachmint Page 3 Pressure-temperature limits for the reactor vessel are developed from the for the most limiting reactor vessel materials. This adjusted adjusted RT ,, determined for a reactor vessel fluence which is bounding for the mustbe RT,S,d per 6 of applicability of the pressure-temperature curves. This is done in accordance with methods from Regulatory Guide No. 1.99, Rev 2, " Radiation Embrittlement of Reactor Vessel Materials." This predicted adjusted RT is used to determine the pressure-temperature limit curves in accordance viUn'the methodology given in B&W Topical Report BAV-10046A Revision 2, " Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G."

License Condition 2C(3)(d) requires that the Toledo Edison Company provide the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection against lov temperature reactor coolant system (RCS) over-pressure events. The analyses described here on the effects of neutron irradia-tion on the RT and the new pressure-temperature curves applicable to 21 EFPY satisfy this r,q,[iirement.

e No plant modifications are necessary to extend lov temperature overpressure protection to 21 EFPY.

Adjusted Reference Temperature Regulatory Guide 1.99, Revision 2 provides guidance on acceptable methods that can be used in calculating the adjusted RT based material properties and the reactor vessel"Eluence. on the Position reactor 2 of vessel this Regula-tory Guide allows the use of surveillance data when two or more credible sets of data are available. Based on the maximum shift of the RT the middle circum-ferenceveldmaterialVF-182-1wasselectedasthelimitinfm,aterial. BAV-2108 estimated the peak fluence at the surfgce at the end of of,this veld material,32-1229012-00 21 EFPY of operation, to be 7.25 x 10 n/cm . B&W calculation 1

used the surveillance data from the four Davis-Besse capsules TEl-F, TEl-B, TEl-A and TEl-D, to determine the RT shift for this veld material. An adjusted RT of155'FvascalculatEatthe1/4Tlocationoftheveld. The adjusted RT at the 3/4T location is 114'F at 21 EFPY.

Pressure-Temperature Curves Using the methodology documented in BAV-10046A Revision 2, the pressure-tempera-ture limit curves for the closure head region, the outlet nozzle region and the beltline region are documented in B&V 77-1229379-03. The following heatup and cooldown rates are assumed in the development of the curves:

Heatup: Heatup rate equivalent to a 50*F/hr ramp from 70'F to 550'F, limited by a 15'F step change followed by an 18 minute hold.

Cooldown: Cooldown rate equivalent to a 100'F/hr ramp from 550'F to 270*F.

limited by a 15'F step change followed by a 9 minute hold.

Cooldown rate equivalent to 50'F/hr ramp below 270'F, limited by a 15'F step change followed by an 18 minute hold.

Heatup and cooldown rates used in the development of the new pressure-temperature curves are the same heatup and cooldovn rates in the existing 10 EFPY curves.

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'/ Lic2nso Nu2ber NPF-3 S rial Numb 2r 2269 Attachment Page 4 A differential pressure correction of -148 psi is applied to the calculated pressure-temperature limits to account for the differential pressure between the analyzed region of the reactor vessel and the system pressure sensor on the reactor coolant system. The differential pressure correction was conservatively estimated assuming that four reactor coolant pumps are operating with an RCS fluid temperature of 200*F.

The maximum allowable pressure (corrected to sensor location) as a function of fluid temperature is determined through a point-by-point comparison for each of the limiting three regions (closure head, outlet nozzle and beltline) stated above. The maximum pressure was taken to be the lowest of the calculated pressures. The resulting data points determined the bounding Technical Specification pressure-temperature curves.

Lov temperature overpressure protection is provided to the RCS in Modes 4 and 5 by the Decay Heat Removal System 4-inch relief valve DH4849 in the DHR System suction line. The DHR system is put in service very soon after entering Mode 4 at 280*F. The relief valve is sized such that it can fully relieve any over-pressure transient which could occur during shutdown. These transients consist of inadvertent operation of the High Pressure Injection System (HPI) and the makeup control valve (MU32) failing open with continuous, makeup pump operation for 10 minutes without operator action. The setpoint of the relief valve is 330 psig. This is less than the allovable pressure limit of 360 psig at 140'F and, therefore, provides adequate protection throughout Modes 4 and 5. The Makeup System is procedurally shutdown at RCS temperatures below 140'F and can no longer impose an overpressure transient on the RCS. Inadvertent initiation of HPI is not considered credible since it vould require multiple failures within ,

the Safety Features Actuation System.

In the event that valve DH4849 is inoperable in Modes 4 and 5, the Action statements for Technical Specification 3.4.2 specify that the HPI pumps be disabled and the total amount of water which can be injected into the RCS by the Makeup System be limited by: (1) restricting the makeup tank level to < 73",

and (2) disabling 'he automatic switchover of the makeup pump suction to the Borated Vater Storage Tank (BVST) on lov makeup tank level. Additional opera-ting restrictions are also provided in Technical Specification figures 3.4-2a and 3.4-2b for operation in Modes 4 and 5, respectively. These figures provide the maximum allowable hot leg pressure as a function of pressurizer level which will allow the RCS pressure to remain bounded by the pressure-temperature limits in the event of a makeup system transient. Because of the conservatism used in the calculation for 10 EFPY and the small change of the pressure limits, it was found that Technical Specification figures 3.4-2a and 3.4-2b remain bounding for pressurization transients through 21 EFPY. The previous calculation did not limit the water injected into the RCS to the volume available in the makeup tank with 73" indicated level. The calculation assumed makeup system injection for the full 10 minutes. This amount of water exceeded the volume available in the makeup tank and resulted in a much greater RCS pressure increase.

For a limited region in Mode 3, administrative controls provide the primary means of maintaining the plant within the limits of the applicable pressure-temperature curve. The potential cause of a lov temperature overpressure event

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Attachment Page 5 in this region is the inadvertent initiation of HPI flow into the RCS or failure of the makeup control valve to a full open position. Inadvertent initiation of  ;

HPI is not a credible event since it would require multiple failures within the Safety Features Actuation System to both establish flow and start the HPI pumps.

Administrative controls vill require the reactor operator to maintain the RCS pressure and pressurizer level within the region where, should the Makeup System control valve fail open, the operator would have ten minutes to respond without exceeding the pressure limitations of the RCS. This situation only exists until the RCS temperature is high enough for the pilot operated relief valve to pro-vide automatic overpressure protection.

Based on the above discussion, it is concluded that the proposed change to Technical Specification Figures 3.4-2, 3.4-3 and 3.4-4, and Bases Section 3/4.4.9, and the extension of the applicability of License Condition 2C(3)(d) from 10 EFPY to 21 EFPY and extending the applicability of Technical Specifica-tion figures 3.4-2a and 3.4-2b without change does not have an adverse effect on safety.

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Attachment Page 6 SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10CFR50.92(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 signifi-cant hazards consideration if operation of the facility in accordance with the proposed changes vould: (1) Not involve a significant increase in the probabil-ity 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 had reviewed the proposed change and determined that a significant hazards consideration does not exist because operation of Davis-Besse Nuclear Power Station, Unit 1, in accordance with this change would:

la. Not involve a significant increase in the probability of an accident '

previously evaluated because: (1) revision of the pressure-temperature curves and the extended applicability of the pressurizer level /RCS pressure limit curves for periods when relief valve DH4849 is inoper-able vill continue to provide the same level of protection of the RCPB as was previously evaluated, and (2) the revision to License Condition '

2.C(3)(d) is administrative to reflect the validi'y of the present analyses to 21 EFPY and (3) the revision to the Technical Specification Bases to reflect the extension to 21 EFPY is administrative and does not affect any previously analyzed accidents.

Ib. Not involve a significant increase in the consequences of an accident previously evaluated because: (1) revision of the pressura-temperature

' curves and the extended applicability of the pressurirer level /RCS pressure limit curves for periods when relief valve DH4849 is inoper- '

able vill continue to provide the same level of protection of the RCPB as was previously evaluated, and (2) the revision to License condition 2.C(3)(d) is administrative to reflect the validity of the present analyses to 21 EFPY and (3) the revision to the Technical Specification Bases to reflect the extension to 21 EFPY is administrative and does not affect any previously analyzed accidents.

2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because (1) revision of the pres-sure-temperature curves and the extended applicability of the pressuri-zer level /RCS pressure limit curves vill continue to provide protection against reactor vessel failure due to brittle fracture concerns under all postulated circumstances, and (2) the revision to License Condition 7.C(3)(d) is administrative to reflect the validity of the present analys>s to 21 EFPY and (3) the revision to the Technical Specification Bases to reflect the extension to 21 EFPY is an administrative change and does ne* affect any activities or equipment in plant operation.

3. Not involve a significant reduction in a margin of safety because:

(1) revisfon of the pressure-temperature curves and the extended applicability of the pressurizer level /RCS pressure limit curves maintains the present margin of safety from reactor vessel brittle fracture as required by 10 CFR 50, Appendix G, and (2) the revision

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'c Liccn32 Nunbar NPF-3 S: rial Nunbar 2269 Attachment Page 7 to License condition 2.C(3)(d) and the Bases revision are administr-ative and do not affect any analyses which provide the basis for the Technical Specifications.

CONCLUSION On the basis of the above, Toledo Edison has determined that the License Amend-ment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifi-cations and a proposed change to the Operating License that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.

ATTACHMENT Attached are the proposed marked-up changes to the Operating License.

References

1. B&W Calculation 32-1229012-00, Adjusted RT "## for 21 EFPY for Davis-Besse I Unit 1, December 1993

2. B&V Calculation 32-1229349-00, Davis-Besse Unit 1 Uncorrected P/T Limits at i 21 FFPY, January 1994

3. B&W Document 77-1229379-03, Pressure-Temperature Limits for 21 EFPY for Davis-Besse Unit 1 Nuclear Power Plant, January 1995

4. BAV-10046A, Rev 2, Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, June 1986

5. BAV-2108, Rev 1, Fluence Tracking System, May 1992

6. BAV-2125, Analysis of Capsule TEl-D, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program- ,

December 1990

7. EAV-1882, Analysis of Capsule TEl-A, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program- , June 1989

8. BAV-1834, Analysis of Capsule TEl-B, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program- , May 1984  !

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9. BAV-1701, Rev 1, Analyses of Capsule TEl-F, The Toledo Edison Company,

• Davis Besse Nuclear Power Station Unit 1 -- Reactor Vessel Surveillance Program- , August 1982

10. BAV-1543A, Master Integrated Reactor Vessel Material Surveillance Program

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