ML20213E095
| ML20213E095 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/07/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213E059 | List: |
| References | |
| NUDOCS 8705140453 | |
| Download: ML20213E095 (6) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 142TO FACILITY OPERATING LICENSE NO. DPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
By a letter dated October 31, 1986, the Iowa. Electric Light and Power Company (the licensee /IELP) submitted an application to amend the Duane Arnold Energy Center (DAEC) Technical Specifications (TSs).
In response to staff questions, the licensee also submitted clarifying information in a letter dated March 20, 1987..The changes were proposed to support the DAEC reload and operation for Cycle 9, and to incorporate administrative changes reflecting revision to figure numbers, table of contents, references, and correction of errors.
During Cycle 9, the licensee proposes to utilize the latest General Electric fuel design (GE88) and analytical methods for fuel analysis (SAFER /GSTR LOCA models and Gemini Physics).
The GE8B fuel design and the improved analytical methods have been previously approved by the staff.
The licensee also proposes to change the TSs by updating the fuel thermal limits of TS Section 3.12, revising ~the Limiting Conditions for Operation and Surveillance Requirements for the Rod Sequence Control System (RSCS) and Rod Worth Minimizer (RWM) in Sections 3.3.B.3 and 4.3.B.3 and modifying the Section 5.2 description of the control blades.
2.0 EVALUATION j
The staff review of the licensee's October 31, 1986, submittal and i
subsequent March 20, 1987, clarifying information is summarized as follows:
Fuel Mechanical Design
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For Cycle 9, 128 irradiated fuel assemblies will be removed from the reactor core and replaced by General Electric 8x8E assemblies.
The i
GE8x8E fuel is similar to that customarily used for 8WR reloads and is I
described in Reference 3.
The mechanical design methodology is described in Reference 5 and was used in this design for the GE8x8E fuel.
Reference l
5 has been approved by the staff in Reference 6 and its supplements.
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2 Fuel Design The new fuel for Cycle 8 is the GE extended burnup fuel GE8x8E.
The fuel designations are BD299A and 80303A.
This fuel type has been approved in the Safety Evaluation Report for Amendment 10 to GESTAR II (Ref. 6).
The specific descriptions of this fuel have been submitted in Amendment 18 to GESTAR II, but since this amendment has not as yet been accepted, the fuel description has also been presented for DAEC in Reference 4.
The fuel descriptions in Reference 4 are acceptable to the staff.
In operation, the GE8x8E fuel will be assigned a number of axial lattice regions and appropriate maximum average planar linear heat generation rate (MAPLHGR) limits.
The MAPLHGR limits have been determined by approved thermal-mechanical and loss-of-coolant accident (LOCA) analyses calculations and will be applied to each of these regions.
There was extensive interaction between the staff, GE and the utility in determining an acceptable format for presentation of this information which is suitable for plant use and meeting staff requirements for TSs.
References 7, 8 and 9 provide questions, responses and conclusions from these interactions.
The process computer contains, and acts on, full details of the MAPLHGR information.
The agreed upon TSs present the least and most limiting lattice MAPLHGR as a function of burnup. When hand calculations of MAPLHGR are required (process computer interactive), the most limiting values are used for all limits.
These TSs are acceptable.
A proprietary report, reviewed by the staff, available to the DAEC engineering staff, provides complete details of the lattice definitions and MAPLHGR limits.
The proposed linear heat generation rate (LHGR) limit for the GE8x8E fuel is 14.4 kW/ft (rather than the 13.4 for other GE fuel).
This LHGR limit has been reviewed and accepted for this fuel in the GE extended burnup fuel review (Reference 10).
This LHGR limit is acceptable for DAEC Cycle 9 operation.
Nuclear Desian The nuclear design and analysis of the Cycle 9 reload was performed with methods and techniques which are described in Reference 5.
The results of the analyses are given in References 3 and 4.
The results of the Duane Arnold analyses are within the range of those reload cores previously reviewed by the staff and found to be acceptable. We therefore conclude that the nuclear design and analysis of the Cycle 9 reload is acceptable.
Thermal-Hydraulic Design The methods and procedures employed in the thermal-hydraulic (T-H) design and analysis of the Cycle 9 core are described in Reference 5.
The value of 1.07 for the safety limit minimum critical power ratio (MCPR), approved in that reference, is used for Cycle 9.
The methods and procedures used to obtain the operating limit MCPR are those described in' Reference 5, approved in Reference 6 and are acceptable.
3 l
Thermal-Hydraulic Stability The issue regarding thermal-hydraulic stability has been resolved during the staff's review of one loop operation (Ref. 11).
The licensee has changed the TSs which provide operating limits and surveillance requirements for thermal-hydraulic stability.
As a result of its review, the staff has determined that the revised TSs implement the recommendations of GE SIL-380 and are acceptable for both one and two loop operation.
Loss-of-Coolant Accident Analyses The LOCA analyses were performed using the SAFER /GESTR code and the application methodology described in Reference 12.
In Reference 15, the staff has specified the necessary conditions for demonstrating applicability of the SAFER /GESTR methodology.
These conditions are:
1.
Calculation of a sufficient number of plant specific peak cladding temperature (PCT) points based on both nominal input values and Appendix K values to verify the shape of the PCT curves versus break size.
2.
Confirmation that plant specific operating parameters have been bounded by the models and inputs used in the generic calculations.
3.
Confirmation that the plant specific emergency core cooling system (ECCS) configuration is consistent with the referenced plant class ECCS configuration.
The licensee has reported the results of those analyses (Ref.' 4) which are required to meet these conditions.
Specifically, the analyses include break sizes from 0.1 ft2 to 2.52 ft2 (DBA recirculation suction line break).
Seven different break sizes were analyzed in conjunction with ECCS failure combinations.
A total of 16 cases were evaluated to establish the trend of PCT curves (nominal and Appendix K) versus break size.
The input parameters for both the nominal and Appendix K cases are within those used in the approved generic analyses.
The ECCS configuration of Duane Arnold (4 LPCI, 2CS and 1 HPCI) is consistent with the ECCS configuration of a generic BWR 4.
The results show that the design basis i
accident (DBA) recirculation suction line break with battery failure is the limiting case.
The calculated PCT is 1036 F when nominal input L
s values are used and 1565* F when Appendix K input values are used.
The input parameters, the ECCS combination and the cases analyzed to establish the trend of PCT verse break size meet the staff requirements given above.
The accident analyses have been performed using approved methods and the results meet the staff's acceptance criteria, therefore, these analyses are acceptable, f
$i MCPR and MAPLHGR Limits A safety limit MCPR has been imposed to assure that 99.9 percent of the, fuel rods in the core will not experience boiling transition during normal operation and anticipated operational transients.
As stated previously, the safety limit of 1.07 was used for Cycle 8.
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4 To assure that the fuel cladding integrity safety limit MCPR will'not be violated during any anticipated transient, the most limiting events were i
reanalyzed for this reload (Ref. 3) to determine which events result in the largest reduction in critical power ratio (CPR).
The operating limit j
MCPR was then established by adding the largest reduction factor in the CPR to the safety limit MCPR.
Since acceptable methods-(Ref. 5) have been used, we find the MCPR TS changes to be acceptable.
The MAPLHGR limits specified in the proposed TS changes are less than or equal to the bounding MAPLHGR used in the SAFER /GESTR-LOCA analysis (Ref. 4) and are, therefore, acceptable.
3.0 TECHNICAL SPECIFICATION CHANGES Reload of Cycle 9 6
The TS changes proposed by the licensee reflect the new fuel for Cycle 9.
These' changes include the LHGR limit, MCPR operating limit and the MAPLHGR f
curve for the GE8x8E fuel.
These proposed changes are acceptable since they are based upon approved analytical methods as discussed above.
Revisions to RSCS and RWM Operability Requirements-In Amendment 12 to Reference 5, General Electric proposed that Group Notch plants which elect to change to Banked Position Withdrawal Sequence i
(BPWS) supervised by RWM for the first 50 percent of withdrawal may take credit for the Control Rod Drop Accident (CRDA) statistical analysis 4
i i
approved by the staff for BPWS plants.
This would result in these plants being able to delete CRDA analysis from reload analysis procedures. As a l
result of the review, the staff has concluded that the proposed amendment i
is acceptable.
The staff has also taken the position that plants electing to change to BPWS must provide a submittal'to the NRC indicating that the BPWS patterns will be enforced and that related TSs would be changed as required to so indicate.
i In response to the staff position, the licensee submitted the proposed changes to the TS.
The DAEC, which has a hard-wired Group Notch RSCS, l
will incorporate the Reduced Notch Worth Procedure (RNWP) into the RWM.
The RNWP is a BPWS compatible but more restrictive sequence. The RNWP will reinforce the control rod withdrawal procedure in the range of 1
highest control rod worth (100% to 50% control rod density)..The J
f existing Group Notch' mode of RSCS will continue to reinforce the rod withdrawal procedure in the range from 50% rod density to the low power setpoint of approximately 30% rated power.
Based on these control rod i
withdrawal procedures, the DAEC TS is changed to reflect these new pnocedures.
The staff has reviewed the TS changes to the RSCS and RWM
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4 and finds these changes acceptable.
l Hybrid I Control Blades The IELP proposed to use several new Hybrid I Control Blades for Cycle 9 operation.
The Hybrid I Control Blades contain both hafnium and boron carbide as neutron absorber materials.
The Hybrid I Control Blades were i
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-designed by General Electric and described in Reference 13.
The staff
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has reviewed and approved the Hybrid I Control Blades (Ref. 14).
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Therefore, the proposed use of the Hybrid I Control Blades for the DAEC is acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
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This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no
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significant increase in the amounts, and no significant change.in the
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types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
- i E.0 CONCLUSIONS Based on the review described above, we conclude that the Duane Arnold Energy Center may be loaded and operated for Cycle 9.
This conclusion is based on the following:
i The safety an' lyses have been performed by previously approved i
1.
a methods and procedures; 2.
The Cycle 9 core meets all of the staff's acceptance criteria.
l The staff also concludes that the associated changes to the TSs for l
Cycle 9 operation, RSCS and RWM operability requirements, and Hybrid I l-control blades are acceptable.
~
l We have concluded, based on t'he considerations discussed above, that:
(1) the're is' reasonable assurance that the health and safety of the 1
1 public will not be endangsred by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to t
the common defense and security or to the health and safety of the public.
i
6.0 REFERENCES
1.
Letter, Richard,W. McGaughy (Iowa Electric Light and Power Company) to Harold Denton (NRC), October 31, 1986.
2.
- Letter, Richard W. McGaughy (Iowa Electric Light and Power Company) to: Harold Denton (NRC), March 20, 1987.
,y 3.
' Supplemental Reload ~ Licensing Submittal for Duane Arnold Atomic Energy Center, Unit' 1, Reload 8, General Electric, 23A 4812, September 1986.
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4.
General Electric Company, NEDO-31310P, Duane Arnold Energy Center,
" SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," August 1986.
5.
GESTAR II
" General Electric Standard Application for Reactor Fuel,"
NEDE-24011-P-A-6, July 1986.
6.
Approval latter, D. G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978, and supplements thereto, forming Appendix C to Referer.ce 5.
7.
NEDE-24081-P, Supplement 1 (and Eratta Sheet No. 11), November 1986,
Loss-of-Coolant Accident Analysis for Peach 8ottom 2," Revision 1.
8.
Letter (and attachments) from J. Gallagher, Philadelphia Electric Co.),
to D. Muller (NRC) dated March 24, 1987, " Peach Bottom 2 Reload 7."
9.
Letter from J. Charnley (GE) to W. Hodges (NRC) dated March 4, 1987,
" Recommended MAPLHGR Technical Specifications for Multiple Lattice Fuel Designs."
10.
Letter (and attachments) from' C. Thomas (NRC) to J. Charnley (GE) dated May 28, 1985, " Acceptance for_ Referencing of Licensing Topical Report NEDE-24011-P-8, Amendment 10."
- 11. ' Letter, A. Thadani (NRC) to L. Liu (Iowa Electric Light and Power Company), May 28, 1985.
12.
NEDE-23785-1-PA, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident" Volume I, II and III, General Electric Company, June 1984.
13.
NEDE-22290, " Safety Evaluation of the General Electric Hybrid I Control Rod Assembly," September 1983.
- 14. Approval letter, Cecil 0. Thomas (NRC) to J. F. Klapproth (GE),
" Acceptance for Referencing of Licensing Topical Report NEDE-22290, Safety Evaluation of the General Electric Hybrid I Control Rod Assembly," August 1983.
15.
Approval letter, Cecil 0. Thomas (NRC) to J. F. Quirk (GE),
" Acceptance for Referencing of Licensing Topical Reports NEDE-23785, Revision 1, Volume III (P), The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident."
Dated: May 7, 1987 r...
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