ML20128L468
| ML20128L468 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/05/1993 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20128L472 | List: |
| References | |
| NUDOCS 9302190260 | |
| Download: ML20128L468 (12) | |
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DUKEPOWER Febmary 5,1993 U. S. Nuclear Regulatory Commission
. ATTN: Document Control Desk Washington, D. C.
20555
Subject:
Catawba Nuclear Station Docket Nos. 50-413 and 50-414 Technical Specification Amendment Supplement Unit 2 Cycle 6 Reload In a letter to the NRC dated December 15, 1992, Catawba Nuclear Station submitt>:d the -
Technical Specification changes that would be necessary for the operation of Catawl,a Unit 2 Cycle 6. Subsequent discussions with the NRC Staff have indicated that the original submittal needed to be supplemented with the following:
1.
A revised no significant hazards analysis which replaces the original no significant hazards analysis in its entirety (Attachment I).
2.
A mark-up of Technical Specification :' l.1. (Attachmem H).
3.
A revised Table 8-1 on page 8-3 of the reload report that changes decreased F n to l
i increased Fan (Attachment II).
In addition, the original submittal used 3.3.3.12 as the technical specification number for the Boton Dilution Mitigation System. Since that time, Amendments 103/97 changed the i
technical specification number for the Baron Dilution Mitigation System to 3.3.3.11.
Accordingly, any reference to Technical Specilication 3.3.3.12 in the original submittal should be changed to 3.3.3.11.
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The original submittal also contained an administrative change to Technical Specification 6.9.1.9 " Core Operating Limits Report" which added a recently approved topical (DPC-NE-1004A) to those already referenced.- Although none of the current changes are based on this topical, the addition _of this topical is necessary to allow future reloads to be carried out under 10 CFR 50.59. If the topical was not added at this time, a future technical
. specification change would be necessary for the sole purpose of adding this topical. 3p fS0"%0$$2!!8886 gd/:E PDR l
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. 4 U..S Nuclear Regulatory Commission February 5,1993 Page 2 If there are any further questions et comments, please contact Chuck Lewis at (803) 831-
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3076.
.i Very truly yours, K. h. \\
M. S. Tuckman Attachments CRUC2C6SUPP.1 1
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t lU.- S. Nuclear Regulatory Commission February 5,1993 -
Page 4
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M. S. Tuckman, being duly sworn, states that he is Vice President, Catawba Nuclear -
Station; that he is authorized on the pan of said company to sign and file with the Nuclear Regulatory Commission this revision to the Catawba Nuclear Station Technical Specifications Appendix A to License Nos. NPF 35 and NPF-52; and that all statements -
and matters set forth therein are tnie and correct to the best of his knowledge.
K.h.k M. S. Tuckman Subscribed and sworn to before me this fu day of Fe 6__,1993.
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+r A1 Npfary Public /
My Commission Expires:
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-U. S. Nuclear Regulatory Commission Fehmary 5,1993.'
- Page 5 bxc: R. C. Futrell G. 'A. Copp G. B. Swindlehurst hi. E. Carroll G.P.Home T. hl. VanDeven J. S. Forbes A. S. Bhatnager S. L. Bradshaw R. bl. Glover S. R. Frye C. R. Lewis NCh1PA-1 NCEh!C Ph1PA SREC Group File: CN-801.01 hiaster File (801.01) 7
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t ATTACHMENTI Revised No Significant Unards Analysis l
NO SIGNIFICANT IIAZARDS ANALYSIS-The following analysis, required by 10 CFR 50.91, concludes that the proposed amendment will not involve significant hazards consideration as defined by 10 CFR 50.92.
10 CFR 50.92 states that a proposed amendment involves no significant hazards consideration if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Create the possibility of a new or different kind of accident from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
POWER DISTRIBUTION AND SAFETY LIMITS Catawba Unit 1 Cycle 6 was the first Duke Power Nuclear Station for which B&W Fuel Company (BWFC) supplied the reload fuel. The Catawba Unit 1, Cycle 6 Reload Report presented an evaluation that concluded tite core reload using Mark-BW fuel would not adversely impact the safety of the plant. The Catawba Unit 1, Cycle 7 report was similar, but reflected that Duke Power perfonned the analyses in support of the operation of Cycle 7 rather than BWFC. This reload for Catawba Unit 2, Cycle 6 is a compilation of the changes made for Unit I during Cycles 6 and 7 in that it justifies the use of Mark BW fuel using Duke Power analysis.
The Catawba Unit 2, Cycle 6 Reload Safety Evaluation Report presents an evaluation which demonstmtes that the core reload using Mark-BW fuel will not adversely impact the safety of the plant. During Cycle 6, the core will contain 76 fresh fuel assemblies supplied by B&W and i17 Westinghouse supplied Optimized Fuel Assemblies (OFAL The changes to the Safety Limit and Power Distribution Technical Specifications presented in Section 8 of the Reload Report represent the application of previously approved methodology to Catawba Unit 2. The changes to remove the power range neutron flux negative rate reactor trip, increase the low steam line pressure setpoint, increase feedwater q
isolation response time, increase steam line isolation response time, increase pressurizer safety valve lift setpoint tolemnce, remove steam line pressure dynamic compensation, increase pressurizer safety valve lift setpoint tolerance, and increase main steam line isolation valve stroke time reflect the use of Duke analysis, and have already been approved for Catawba Unit 1. The changes described above include the deletion of references to specific units on individual Technical Specification pages, and delete pages which were I
4 previously for Unit 2 only. The implementation of unit' specific references became necessary due to the transition from Westinghouse to B&W supplied fuel during Unit 1 Cycle-6 and for the Unit 1 Cycle 7 Reload due to the transition to Duke analysis methodology. The analysis which made' the changes necessary in.the Unit:I reload submittal is generic, and as described in the technicaljustification, is equally applicable to -
i both hicGuire and Catawba units.
A LOCA evalua s a for operation of Catawba Nuclear Station with Mark-BW fuel has been completed (BAW 10174, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units). Operation of the station while in transition from Westinghouse supplied OFA fuel -
to B&W supplied Mark-BW fuel is also justified in this topical.
BAW-10174 demonstrates that Catawba Nuclear Station continuu to meet the criteria of-10 CFR 50.46 when operated with Mark-BW fuel. Utrge Break LOCA calculations completed consistent with an approved evaluation model (BAW-10168P and revisions) demonstrate compliance with 10 CFR 50.46 for breaks up to and including the double ended severance of the largest primary coolant pipe. The small break LOCA calculations ;
used to license the plant during previous fuel cycles are shown to be bounding with respect to the new fuel design. This demonstrates that the plant meets 10 CFP 50.46 criteria when the core is loaded with Mark-BW fuel.
During the transition from Westinghouse OFA fuel to Mark-BW fuel, both types of fuel:
assemblies will reside in the core for several fuel cycles. Appendix' A to BAW-10174 demonstrates that results presented above apply to the Mark-BW fuel in the transition core, and that insertion of the Mark-BW fuel will not have an adverse impact on the cooling of 1
the Westinghouse fuel assemblies.
Duke Power Company's Topical Repons DPC-NE-3000, DPC-NB-3001-PA, and DPC-NE-2004-PA provide evaluations and analyses for non-LOCA transients which are applicable to Catawba. The scope of these analyses includes all events specified by sections 15.1-15.6 of Regulatory Guide 1.70 (Standard Fonnat and Content of Safety Analysis Reports for Nuclear Power Plants) and presented in the Final Safety Analysis Report for Catawba. The analysis and evaluations perfonned for these topicals confinn that operation of Catawba Nuclear Station for reload cycles with Mark-BW fuel will continue to be within the previously reviewed and licensed safety limits.
1 One of the primary chjectives of the Mark-BW replacement fuel is compatibility with the resident Westinghouse fuel assemblies. The description of the Mark-BW fuel design and the thennal hydraulics and the core physics perfonnance evaluation demonstrate the-similarity between the reload fuel and the resident fuel. The extensive testing and analysis summarized in BAW-10173P shows that the Mark-BW fuel design perfonns, from the standpoint of neutronics and thermal-hydraulics, within the bounds and limiting design criteria applied to the resident Westinghouse fuel for the Catawba plant safety analysis.
2
Each FSAR accident has been reviewed to detennine the effects of Cycle 6 operation and to ensure that the radiological consequences of postu'2ted accidents are within applicable regulatory guidelines, and do not adversely affect the health and safety of the public. The design basis LOCA evaluations assessed the radiological impact of differences between the Mark-BW fuel and Westinghouse OFA fuel fission product core inventories. Also, the dose calculation effects from non-LOCA transients reanalyzed by Duke Power were-evaluated using Cycle 6 characteristics. The calculated radiological consequences are all within specified regulatory guidelines and contain significant _ levels of margin.
The analyses contained in the referenced Topical Repons indicate that the existing design criteria will continue to be met. Therefore, the enclosed TS changes will not increue the probability or consequences of an accident previously evaluated.
As stated in the above discussion, nonnal operational conditions and all fuel related transients have been evaluated for the use of Mark-BW fuel at Catawba Nuclear Station.
Testing and analysis was also completed to ensure that, from the standpoint of neutronics and thermal-hydraulics, the Mark-BW fuel would perfonn within the limiting design criteria. Because the Mark-BW fuel perfonns within the previously licensed safety limits, the possibility of a new or different accident from any previously evaluated is not created The reload-related changes to the TSs do not involve a signincant reduction in the margin of safety. The calculations and evaluations documented in BAW-10174 show that Catawba will continue to meet the criteria of 10 CFR 50.46 when operated with Mark-BW fuel.
The evaluation of non-LOCA transients documented in DPC-NE-3001 also confinns that Catawba will continue to operate within previously reviewed and licensed safety limits.
Because of this, the TS changes to suppon the use of Mark-BW fuel will not involve a significant reduction in the margin of safety.
The technical changes made to Table 2.2-1 reflect the use of the BWCMV CIIF correlation and Duke Power's Statistical Core Design methodology with a 1.55 thennal design limit.
These changes to Table 2.2-1 will not significantly increase the probability or consequences of an accident previously evaluated. the changes to the K values conservatively bound the allowable operating region, as deOned by the new DNBR methodology.
It can be
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concluded that these changes will not create the possibility of any new accident from those previously evaluated. It can also be concluded that since all new TS values are bounded by safety analysis assumptions that this change will not significantly decrease the margin of safety.
.i DELETION OF NEUTRON FLUX IIIGII NEGATIVE RATE TRIP The removal of the Power Range Neutron Flux High Negative Rate trip will not result in any previously-reviewed accident becoming more probable or more severe. The trip is a response to a pre-existing transient condition and would not initiate any accident. The trip 3
is designed to provide protection from a dropped control rod. Ilowever, in the_ event of a dropped rod, the reactor is assumed to trip on low pressurizer pressure. - Therefore, the -
protection function is retained. The consequences of a dropped rod have been analyzed and -
found to be within acceptable limits.
Likewise, the removal of this trip will not create a new accident not previously reviewed.'
The temoval of a response to a transient will not initiate a new transient. There are no credible unanalyzed transients which will occur as a result of a dropped rod. The removal of this trip will reduce the potential for spurious or unnecessary trips which may occur as a result of maintenance or the drop of a low-worth rod There are no other hardware modifications or procedum changes that will be made as a result of this deletion which could create the possibility of a new accident.
No margin of safety will be reduced by this change. As noted above, if a dropped rod necessitates a trip, the trip function will be accomplished as a result of low pressurizer pressure. For those dropped rods for which no trip is necessary, the removal of this trip will provide protection against an unnecessary transient.
LOW STEAM LINE SETPOINT PRESSURE CIIANGE Changing the Low Steam Line Pressure setpoint and removal of dynamic compensation will not increase the probability or consequences of any previously-reviewed accident _ The higher steam line pressure setpoint is consistent with all licensing basis safety analyses.
This change, in conjunction with the removal of the dynamic compensation of the steam pressure signal, is intended to reduce or climinate spurious Engineered Safeguards Features (ESF) actuations which are caused by minor (but rapid) pressure decreases in the secondary system.
The proposed amendment will not result in a new accident not previously reviewed. A change in steam line pressure is a response to an existing transient condit!cn, rather than a precursor or initiating event. A change in the steam line pressure setpoint is also not a precursor or initiating event.
The proposed amendment will not result in a significant decrease in a margin of safety.
The reanalysis of the steam line break accident which was performed shows that all imposed Condition II acceptance criteria are met.
Based on the above, it is concluded that ne significant hazards exist.
4
s FEEDWATER AND MAIN - STEAM LINE-ISOLATR)N VALVE STROKE TIME The proixised changes to the valve stroke times in Tables 3.3-5 and Table 3.6-2a will not significantly increase the probability 01 consequences of any previously evahiated accident.
The effects of the delays in isolation times on the various transients affected have been analyzed and found to be acceptable. Since these valves do not receive a containment isolation signal, and no credit is taken for operation of these valves in the dose analysis for a containment isolation function, a maximum stroke time does not apply for containment isolation.
The proposed changes will not significantly increase the possibility of a new accident not previously evahiated.
Feedwater and main steam isolation are responses to ongoing transients, rather than initiators or precursors of tmnsients. No equipment or component reconfiguration will occur as a result of this change.
The proposed changes will not significantly decrease any margin of safety. As noted above, the effects of the longer isolation times have been evaluated and found to be acceptable.
Based on the above, it is concluded that no significant hazards exist.
INCREASE IN PRESSURIZER CODE SAFETY VALVE SETPOINT TOLERANCES The proposed amendment will not result in a significant increase in the probability or consequences of any previously analyzed accident. The valve lift setting is challenged only after a transient has been initiated and is not a contributor to the probability of any transient or accident. The transients which involve pressure increases which would potentially challenge the safety valves have been analyzed to deterinine the consequences of delayed or premature valve actuation at the extremes of the new setpoint tolerances. These analyses show that all applicable acceptance criteria are met using the wider tolerances.
The proposed amendment will not result in the creation of any new accident not previously evaluated. As noted above, the setpoint tolerance only affects the time at which the safety valve opens following or during a transient, and is not a contributor to the probability of an accident.
The proposed amendment will not result in a significant decrease in a margin of safety.
The limiting transient in each accident category has been analyzed to determine the effect 5
of the change in lift setpoint tolerance on the tmnsient. In each case, the results of the analyses met all acceptance criteria.
Based on the above, it is concluded that no significant hazards exist.
CONTAINMEN'I' ISOLATION VAINES The proposed changes to the valve stmke times in Table 3.6-2a and 3.6.2b will not significantly increase the probability or consequences of any previously evaluated accident.
The effects of the delays in isolation times on the various transients affected have been analyzed and found to be acceptable. Since these valves do not receive a containment isolation signal, and no credit is taken for operation of these valves in the dose analysis for a containment isolation function, a maximum stmke time does not apply for containment isolation.
The proposed changes will not significantly increase the possibility of a new accident not previously evaluated.
Feedwater and main steam isolation are responses to ongoing transients, rather than initiators or precursors of tmnsients. No equipment or component recon 0guration will occur as a result of this change.
The proposed changes will not significantly decrease any margin of safety. The isolation times which are applicable to these valves are specified in Table 3.3-5, Engineered Safety Features Response Times. The effects of the isolation of these valves was evaluated based-on their ESF function, not a containment isolation function, and determined to be acceptable, therefore there is no significant decrease in the margin of safety.
BORON DILUTION MITIGATION SYSTEM TS 3.3.3.11.a.2 is changed to reduce the allowable Reactor hiakeup Water Pump flow in hiode 5 from 75 gpm to 70 gpm. In the event that the Boron Dilution Mitigation System (BDMS) is inoperable the Reactor Makeup Water Pump flowrates are limited to ensure that operator action times required to terminate a dilution event can be met. The limits on reactor makeup water pump Dowrates when the BDMS is inoperable are verified each cycle i
to ensure that the safety analysis assumptions for these parameters remain valid. When the l
calculated Reactor Makeup Water Pump flowrate is found to be less than the existing =
flowrate limits, the Dowrate limit _must be reduced so that the operator action time acceptance criteria of Standard Review Plan 15.4.6 can be met.
Reducing the allowable Reactor Makeup Water Pump flow in Mode 5 does not involve a significant increase in the probability'or consequences of an accident previously eva!uated.
The current TS Howrate does not allow enough time for the operator to tenninate an I
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uncontrolled dilution event when required operator response times are assumed. The lower flowrate allows needed operator response times and is therefore more conservative.
Reducing the allowahle Reactm hiakeup Water Pump flow in hiode 5 does not change tiie way that any plant equipment is operated or maintained, thervfore it does not create the.
possibility of a new or different accident.
Reducing the Allowable Reactor hiateup Wat:r Pmnp Flow in hiode 5 will not involve a significant reduction in the margin of safety..This flowmte is more conservative, and ensures that safety analysis assumptions regarding operator actions times in response to the tennination of an uncontrolled dilution event can be met.
CORE OPERATING LIhllTS RI? PORT The proposed change to TS 6.9.1.9 adds approved topical DPC-NE-1004A to the list of analytical methods used to detennine core operating limits. This chang:is administrative, adding a topical report which has been approved for use on Catawba to the list of analytical methods used to detennine core operating limits. Since this change is administrative it has been detennined that no significant hazards are involved.
ENVIRONh1 ENTAL IMPACT STATEh1ENT The proposed Technical Specification change has been reviewed against the criteria of 10-CFR 51.22 for environmental considerations. As shown above, the proposed change does not involve any significant hazards consideration, nor increase the types and amounts of effluents - that may be released offsite, nor increase the individual or cumulative occupational radiation exposures. Based on this, the proposed Technical Specification change meets the criteria given in 10 CFR 51.22 (c) (9) for categorical exclusion from the requirement for an Environmental Impact Statement.
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