ML20023C149
| ML20023C149 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/25/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20023C148 | List: |
| References | |
| NUDOCS 8305110560 | |
| Download: ML20023C149 (7) | |
Text
9 Katuqk UNITED STATES
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NUdLEAR REGULATORY COMMISSION 5 *f j
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 80 TO LICENSE NO. DPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY 2
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET N0. 50-331 DUANE ARNOLD ENERGY CENTER 1.0 Introduction By letter dated January 14,1983 (Bef.1), the Iowa Electric Light & Power Company (licensee) made application to amend the Technical Specifications of Operating License DPR-49 for the Duane Arnold Energy Center (DAEC) in order to operate the plant for fuel Cycle 7.
In support of this application, the licensee also provided a supplemental reload licensing submittal (Ref. 2).
We have reviewed these submittals and'oEE evaluation follows.
2.0 Evaluation 2.1 Fuel Desian Evaluation The reload application contains four fuel-design related issues:
(.1) the replacement of all 7X7 fuel assemblies with the newer P8X8R fuel assemblies, (2) the analysis of safety considerations involved in the determination of Cycle 7 operating limits, (3) the reanalysis of the loss-of-coolant accident (LOCA) with the incorporation of extended maximum average planar linear heat generation rate (MAPLHGR) limits, and (4) the reanalysis for the control rod d' rop accident (RDA).
Replacement of 7X7 Fuel Assemblies The Cycle 7 reload fuel is comprised of 128 standard-design P8X8R fuel assemblies.
These assemblies will replace the last of the 7X7 prede-cessor fuel assemblies.
The' Cycle 7 core inventory is given in Table 1.
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- TABLE 1 DUANE ARNOLD CYCLE 7 CORE INVENTORY Assembly Designation Cycle Loaded Number
- S 8DB274H 4
68 P8DB289 5
88 P8DPB289 6
84 P8DRB284H 7
88 P8DRB299 7
40 368
- All assemblies are drilled.
Cycle 7 Operating Limits The licensee's analysis of the safety considerations involved in the
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determination of Cycle 7 operating limits is set forth in the reload report (Ref. 2).
In all fuel-design-related area 7, except those sepa-
'C rately identified, the reload report relies on the generic report, General Electric Standard Application for Reactor Fuel (Ref. 3), which we previously reviewed and approved (Ref.14).
Loss-of-Coolant Accident The licensee has submitted (Ref. 4) the results of a LOCA analysis that addresses the fresh Cycle 7 fuel and revised MAPLHGR limits for all Cycle 7 fuel types.
The LOCA analysis was performed using th'e General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K (Ref. 5) as amended (Ref. 6) in 1977.
In 1981, the NRC conditioned (Ref. 7) the use of the GE emergency core cooling system (ECCS) evaluation model (EH) to require that plant analyses performed with the GE evaluation model be accompanied by supplemental calculations performed with a specified set of material correlations from NUREG-0630
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(Ref. 8).
This condition was subsequently removed (Ref. 9) following a l
GE modificatiion to the cladding rupture temperature model.
The DAEC LOC.A analysis did not utilize the new " adjusted" GE rupture temperature model.
flevertheless, the licensee has addressed (Ref.10) this issue
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and de'termined that for the DAEC Cycle 7 application there would have been no difference in LOCA analysis if tha " adjusted" model had been employed. This is because the ruptures predicted for DAEC.are in the l
temperature regime where the " adjusted" model coincides with the origi-nal GE model.
Consequently, we conclude that the licensee's use of the
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original GE rupture temperature model'is acceptable.
The licensee's submittal also provided revised MAPLHGR limits that have been extended to accommodate an exposure of 45 GWd/MTU.
These limits were generated by methods previously approved (Ref. 6).
Although the methodology used is generically applicable for the MAPLHGR limit determi-nation, we believe that the effects of enhanced fission gas release in
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high-burnup fuel (i.e., greater than 20 GWd/MTU) were not adequately con-i sidered in the fuel performance model.
In response to this concern, GE requested (Refs.11 and 12) that credit for approved, but unapplied, ECCS evaluation model changes and calculated peak cladding temperature margin i
be used to avoid MAPLHGR penalties at higher burnups.
We"found this proposal acceptable (Ref.13),.provided that certain plant-specific conditions were met.
The licensee has stated (Ref.- 10) that the GE proposal is applicable to the Duane Arnold analysis. On this b6 sis, we conclude that the MAPLHGR j
limits proposed for Cycle 7 operation of Duane Arnold are in conformance with the requirements of 10 CFR 50.46 and are acceptable.
Control Rod Droo Accident
.l A reanalysis for the control rod drop accident was performed with. generic-'
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bounding and plant-specific inputs.
The resultant peak enthalpy was'
.I found to be 236.6 cal /g.
The' calculated value is less than the acceptance
criterion (i.e., 280 cal /g) given in Section~ 15 A9 of the Standard Re-view Plan (flVREG-0800).
We, therefore, conclude that the analysis meets the pressure boundary integrity and coolability requirements of the General Design $riterion 28 of Appenoix A to 10 CFR 50 and is, hence, acceptable.
1 Changes to the Technical Specifications 2
Changes to the Technical Specifications, concerning the replacement of 7x7 fuel l
' assemblies, Cycle 7 operating limits, and loss-of-coolant accident are acceptable, i
based on the above evaluation.
2.2 Thermal and Hydraulic Design Evaluation The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable methods, and provides i
an acceptable margin of safety from conditions which could lead to fuel
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damage during normal and anticipated operational transients, and that'the core is not susceptible to thermal-hydraulic in' stability.
a The review includes the following areas:
(1) safety limit minimum critical power ratio (MCPR), (2) opeirating limit MCPR, (3) thermal-hydraulic stability, and (4) changes to Table 3.12-2 of the Technical Specifications.
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The licensee has submitted an analysis report for Cycle 7 operation (Ref.2).
This repor,t relies on a generic document (Ref. 3)'which has" been reviewed and approved (Ref.14) by the staff.
Discussion of our review concerning the thermal-hydraulic, design for Cycle 7 operation-follows.
Safety Limit MCPR A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected -to. experience boiling. transition
.during nonnal and anticipated operational transients.
As stated in Reference 3, the. approved safety limit MCPR is' 1.07.
This safety limit MCPR of.1.07-is used for-the DAEC Cycle 7 operation.
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,.5-Operatino Limit MCPR The most limiting events have been analyzed by the licensee to determine c
which event could potentially induce the largest reduction in the initial critical power ratio (aCPR).
The ACPR values given in Item 9 of
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Reference 2 are plant-specific values calculated by including the ODYN 4
computational method.
The calculated ACPRs are adjusted to reflect Option A ACPRs by employing the conversion method described in Reference 15.
The MCPR values are determined by adding the adjusted aCPRs to the safety limit MCPR.
The maximum cycle MCPR values (Option A) in Item 11 of Re-ference 2 are specified as the operating limit MCPRs and incorporated into the Technical Specifications.
Since an approved method was used to deter-mine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated transients, we conclude that these 7
limits are acceptable.
Thermal-Hydraulic Stability The results of the thermal-hydraulic analysis (Ref. 2) show that the maximum reactor core stability decay rat'io is 0.85, which is less than the calculated value for some operating reactors which have been prevd-ausly approved.
Since the operation in the natural circulation mode will be prohibited by Technical Specification 3.3.E(LCO), there will be added margin to tne stability limit and we, therefore, conclude that the
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thermal-hydraulic stability results are acceptable for Cycle 7 operation.
Chances to the Technical Specifications '
Changes to 'the Technical Specifications concerning the safety limit minimum critical power ratio (fiCPR) and the operating limit'MCPR for Cycle 7 t
are acceptable, based on the above evaluation.
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2.3 Evaluation Summary 1
j We have reviewed the fuel and thermal-hydraulic design.related issues sub-I j
mitted, and we find, based on the above, that the~ reload safety analysis
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for Cycle-7 operation of DAEC, including the necessary changes to the Technical Specifications are: acceptable.
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6-l 3.0 Environmental Consideration We have determined that the amendment does not authorize a change in-effluent types or total amounts nor an increase in power level and will not resul.t in any significant environmenta). impact. -Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the stand environmental impact and, pursuant to 10 CFR 551.5(d)(4) point of that an i
environmental impact' statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusion We have concluded, based on the considera"tions discussed above, that:
(1) because the amendment does not involve a' significant increase in the probability or consequences of an accident previously evaluated, does not create the possibiitty of an accident of a type different from any evaluated previously, and does not involve a significant reduction in.a margin of safety, the amendment does not involve a significant hazards consideration, (2) there is. reasonable assurance that the health and safety of the public' will not be endangered by operation in the proposed manner,,. and (3) such. activities will be conducte'd in compliance with 'the Commission's regulations and the issuance of this ameddment will not be inimical to the common
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defense and security or' to. the health and safety of.the public,.
Dated: April 25, 1983 Principal Contributors:
D. Powers, T. Huang i
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J 5.0 References 1.
L. D. Root (IEL&P) letter to H. R. Denton (USNRC), January 14, 1983.
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2.
" Supplemental Reload Licensing Submittal for Duane Arnold Atomic Energy Center Reload 6," GE Report Y1003J01 A46, January 1983.
3.
' " General Electric Standard Application for Reactor Fuel." GE Report NEDE-240ll-P-A-4, April 1978.
4.
" General Electric, Loss-of-Coolant Accident Analysis Report for l
Duane Arnold Energy Center (Lead Plant)," GE Report NED0-210282 l 1A, Revision 2, June 1982.
5.
" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K," GE Report NED0-20566, 1974.
6.
D. G. Eisenhut (USNRC) letter to E. D. Fuller (GE), June 30, 1977.
7.
R. L. Tedesco (USNRC) letter to G. G. Sherwood (GE), February 4,1981.
1 8.
D. A. Powers and R. O. Meier, " Cladding Swelling and Rupture Models for LOCA Analyses," NRC Report NUREG-0630, April 1980.
9.
H. Bernard (USNRC) lett.er.to G.,G. Sherwood (GE), May 11,1982.
10.
L. D. Root (IEL&P)-letter to H. R. Denton (USNRC), April 7,1983.
11.
R. E. Engel (GE) letter to T. A. -Ippolito (USNRC), May 6,1981.
12.
R. E. Engel (GE) letter to T. A. Ippolito (USNRC), May 28, 1981.
13.
L. S. Rubenstein (USNRC) memorandum for T. M. Novak, " Extension of General Electric Emergency Core Cooling Systems Performance Limits,"
June 25, 1981.
14.
Letter from D. G. Eisenhut (USNRC) to R. Gridley (GE), May 12, 1978.
15.
NEDE-24154, " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors ~" October 1978.
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