ML20028C059
| ML20028C059 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/23/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028C058 | List: |
| References | |
| NUDOCS 8301050493 | |
| Download: ML20028C059 (5) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENOMENT NO.19 TO FACILITY OPERATING LICENSE DPR-77 TENNESSEE VALLEY AUTHORITY INTRODUCTION By letter dated September 17, 1982, the Tennessee Valley Authority (TVA) made appli-cation to amend the Technical Specifications for the Sequoyah Nuclear Plant, Unit 1, in order to reload and operate the plant for Cycle 2.
In support of the appli-cation TVA submitted a reload safety analysis requesting changes in fuel enrich-ment and operating procedures.
A December 13, 1982, letter from TVA transmitted additional infomation regarding themal hydraulic parameters.
DISCUSSION As part of their reload Technical Specification amendment, TVA has proposed to change the U-235 enrichment of the fuel assemblies in the core. This modification is being made to facilitate an eighteen month cycle versus the previous twelve month cycle.
Sixty-eight assemblies of 2.1 weight percent U-235 enrichment representing the initial lowest enrichment of Cycle I will be discharged, and sixty-eight fresh assemblies having a 3.65 weight percent U-235 will be added to the core periphery.
Additionally, the licensee has proposed several changes to the power distribution control procedure, discussed in Tech Spec 3/4 2.1, for Cycle 2.
First, the Relaxed Axial Offset Control Procedure will replace the Constant-Axial Offset Control Procedure.
Second, the presently used surveillance procedure for F (z) is replaced by a revised 9
procedure which facilitates more convenient adherence to this requirement.
Finally,.
the coefficient of the power dependent terni in the F Technical Specifications 2.1,
~2.2, and 3/4 2.3 has been revised to pemit larger vafues of this qualtiy at lower i
power.
These changes are proposed to increase operational flexibility and to permit optimization of the core loading pattern for full power operation.
EVALUATION In evaluating changes incurred when increasing U-235 enrichment, neutronics and thermal. hydraulics parameters were analyzed. - The neutronics parameters for Cycle 2 were obtained through utilization of.the staff-approved PALA00N Code.
PALA00N is a three-dimensional nodal analysis program which is used to calculate core power distributions, reactivity coefficients, rod worths, and other kinetics parameters.
Comparison of these values for both cycles is found in Tables 2 through 4 of TVA's September 17 transmittal. Cycle 2 parameters were bounded by those of Cycle 1 with the exception of the least negative value of the Doppler temperature coefficient (DTC).
This value increased approximately 29% but since the DTC represents a small part of the total negative reactivity feedback, the consequences of the various events were changed only insignificantly.
Consequently, no accidents were reanalyzed. The
. staff finds this acceptable.
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. The required shutdown margin for Sequoyah is 1.60 %
While the actual shutdown margin has decreased between cycles (Table 3 of September 17, 1982 submittal),the Cycle 2 end-of-core-life value is 26% above'the requirement of 1.60 %
This is acceptable to the staff.
Themal hydraulic parameters for both cycles as provided by TVA in their December 13, 1982, transmittal are summarized in the enclosure.
The parameters given show no variations that would affect the thermal mapins as a result of the Cycle 2 reload.
The design reactor coolant flow is 138.0g10 lb/hr (365,600 gpm), whereas the most recently measured flow rate is 144.85X10 lb/hr (383,763 gpm) which indicates an ample margin in flow rate.
The proposed amadment requests three changes to the power distribution control pro-cedure.
In replacing the Constant Axial Offset Control (CAOC) Procedure with the Relaxed Axial Offset Control (RA0C) Procedure, the licensee enhances maneuvering capability, limits operator action required for confomance to power distribution Technical Specifications, and increases return to power capability following a reactor trip.
The RAOC procedure is described in Westinghouse topical report, NS-EPR-2649, Part A, which is referenced in the application.
This report has been reviewed by the staff which concluded that the RAOC procedure is an acceptable method for power distribution control in Westinghouse reactors.
TVA also wishes to substitute a revised surveillance procedure for F (z) for the 0
present one.
The revised surveillance is described in Westinghouse report HS-EPR-2649, Part B.
This report has been reviewed by the staff and found accept-able.
The revision allows the licensee to ensure plant operation below the F (*)
0 limit more easily.
The Fn(z) surveillance has no effect on Cycle 2 analyses and safety parameters.
This change requires that a Peaking Factor Limit Report citing cycle dependent factors be submitted at least 60 days prior to startup.
This reporting should reduce technical specification changes.
By letter dated October 19, 1982, TVA submitted a Peaking Factor Limit Report for Cycle 2 which indicates l
thAt TVA has elected to use the option of breaking the cycle into exposure segments in order to increase operational flexibility.
This is acceptable.
The last item dealing with the power distribution control procedure seeks to revise the coefficient of the power dependent term, K, in the F g Techncial Specification.
l The coefficient will be increased to 0.3 from its present value of 0.2.
This has the effect of increasing the allowed radial peaking factor at low powers.
This, in l
turn, affects the themal hydraulic behavior, rod worths and kinetics parameters.
The effects of these changes have been factored into the reload analysis discussed above.
In particular the hot-zero power, end-of-life rod ejection cases were reanalyzed and still neet the acceptance criteria for this event.
The increased radial peaking factor at lower power is included in the RAOC analysis which determines the allowed axial offset values for the plant and is also used in the deternination of the overtemperature-delta temperature trip algorittn.
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The reactor core safety limits given in Figure 2.1-1 of the proposed Technical Speci-fications have been revised to reflect the increase of K from 0.2 t'o 0.3 in the follow-ing relationship:
F'g _,1.5 [1 + K(1-P)]
where P = fraction of rated power for power levels less than 100L This increase in K has been previously approved for the Trojan application (letter from C. Trammell, NRC, to B. Withers, Portland General Electric, August 13,1982)and is included in the Trojan Technical Specifications.
The staff concludes that the effect of the changed constant in the F H Technical Specification has been properly accounted for and is acceptable.
Another change involves removing an unnecessary statement from Section B 3/4 2.4 describing quadrant power tilt ratio.
The staff finds this change acceptable.
The licensee also requested removal of the interim operating restrictions imposed as a result of deficiencies in the analysis of the control rod drop event. Westinghouse.
has submitted a generic topical report supporting the removal of these restrictions when certain analyses are perfomed. The-review of this report has not been com.
pleted by the NRC staff. Until such time as the review is complete and any required analyses have been performed, interim operating restrictions must remain in effect.
Also included in the September 17, 1982, transmittal were requests to amend Tech-nical Specifications 3.6.1.5 and 3.6.1.9 which change containment air temperature and i
increase the number of purge lines.
Staff evaluation of these changes will be the subject of a later amendment.
ENVIRONMENTAL CONSIDERATION p
We have determined that the amendment does not authorize a change in effluent types I
or total amounts nor an increase in power level and will not result in any signif-icant environmental impact.
Having made this determination, we have further con-cluded that the amendment involves an action which is insignificant from the stand-point of environmental impact and, pursuant to 10 CFR 51.5(d)(4),that'anenviron-mental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that: (1)'because l-the amendment does not involve a significant increase in the probabilityior con-sequences of accidents previously considered, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant decrease in a safety margin, the amendment does not involve a
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significant hazards consideration, (2) there is reasonable assurance that the health omce) l sunn=e >
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and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regu-lations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: December 23, 1982 Principal Contributor: Carl Stahle, Licensing Branch No. 4, DOL Melanie Hiller, Licensing Branch No. 4, DOL Walter Brooks, Core Perfonnance Branch, DSI Harry Balu'kjian, Core Performance Branch, DSI
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4 Thermal Hydraulic Parameters -
Cycle 1 Cycle 2-I.
Perfomance Characteristics:
Total Heat Output Mw(t) 3411 3411~
System Pressure Nominal Psia 2250 2250 Minimum in Steady-State 2200 2200 Coolant Flow:
Total Design Flow Rate 106 lb/hr 138.0 138.0 Total Design Flow Rate gpm 365600 365600 4
Coolgnt Design Flow through Core 10 lb/hr 127.7 127.7.
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Coolant Design Flow through Core gpm 33818')
'l38180'~
Pressure Drop Across Core psi 23.4 1 2.3
.23.4 1 2.3-s; 6
2 Average Mass Velocity 10 lb/hr-f t 2.50
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Coolant Tenperature:
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Design Inlet Temperature F-545.7 546.7 t
Average core Temperature Rise F 67.6 s'
- 67.6
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Heat Transfer:
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ActiveHegtTransferSurface Area ft 59,700
-59,70D t
Core Averaga lleat Flux 1-BTU /hr-ft2 184800 '
18980'O f
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Average Linear Heat Rate (kW/ft) 5.44' 5.44
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,f Maximum Linear lieat Rate (kW/ft)
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,J Peak Linear Heat Rate (LW/ft)
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Ninimtn DNBR at Nominal Conditions p
3 for Typical Flow Channel 2.22 j 2.52,f.
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I Minimun DNBR for Design Transients
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1}Di; for Thimble (cold wall) flow channel 1.81 3
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