ML042260033
| ML042260033 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/03/2003 |
| From: | Duke Energy Corp |
| To: | Office of Nuclear Reactor Regulation |
| Byrdsong A T | |
| References | |
| 50-413-OLA, 50-414-OLA, ASLBP 03-815-03-OLA, Catawba-Applicant-2, RAS 8248 | |
| Download: ML042260033 (39) | |
Text
- USNRG'-
August 9, 2004 (11:45AM)
(
OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF Exhibit 2 Relevant Portions of Duke Energy's Response to NRC Request for Additional Information Submitted to the NRC, November 3, 2003 NUCLEAR REGULATORY COMMISSION Docket No. -4i34tIfqM-OL Official Exh. No. '
In the matter of __ _ _ _ _ _ _ _ _ _ _ _
Staff _ IDENTIFIED 7/N /A' Applicnt __ RECEIVED 7/lf lot.
Intel' :nr - __ -REJECTED Conig 0!'t _
Contiamcto- DATE Other Witness too"gyzea -'
,1,epla-te - SYeC V- oils Secy-og
Exhibit 2
- 12. Provide the appropriate regulatory criteria to be satisfied by the information in section 3.7, i.e., how this section meets the general design criteria specified in the Standard Review Plan.
Response (Previously submitted October 3, 2003)
Section 3.7 contains the safety analysis of three distinct subject areas; loss of coolant accidents (LOCA), non-LOCA accidents, and radiological consequences. The appropriate regulatory criteria for each of these topics are summarized in Tables Q12-1 through Q12-3.
LOCA Criteria The LOCA acceptance criteria of 10CFR 50.46 (b) were established for light water reactors fueled with U0 2 pellets within cylindrical Zircaloy cladding. The MOX fuel lead assemblies have M5w cladding and mixed oxide fuel pellets. The applicability of the 10CFR 50.46 criteria to the MOX fuel lead assemblies is established in Table Q12-1.
Non-LOCA Criteria The criteria used to evaluate the non-LOCA transients/accidents in the Updated Final Safety Analysis Report are summarized in Table Q12-2 and except for rod ejection accident criteria are the same criteria used for analysis of non-LOCA transients/accidents in LEU fuel cores.
Provisional Rod Ejection Accident Criteria The current acceptance criteria for a rod ejection accident (REA) at Catawba are described in Section 4.1.2 of Reference Q12-1. These criteria are based on Section 15.4.8 of the StandardReview Plan (Reference Q12-2), and are summarized below.
- 1. The radially averaged fuel pellet enthalpy shall not exceed 280 cal/gm at any location.
- 2. Doses must be "well within" the 10 CFR 100 dose limits of 25 rem whole-body and 300 rem to the thyroid, where "well within" is interpreted as less than 25% of those values.
- 3. The peak Reactor Coolant System pressure must be within Service Limit C as defined by the ASME Code, which is 3000 psia (120% of the 2500 psia design pressure).
With the exception of the enthalpy limit of 280 calgm, those criteria are equally valid for mixed oxide (MOX) fuel as for low enriched uranium (LEU) fuel during a REA. The dose acceptance criteria relate to the radiological consequences to the public, not the fuel type. The primary system pressure acceptance criterion relates to the integrity of the pressure boundary, not the fuel type.
The enthalpy limit was established to ensure coolability of the core after a REA and to preclude the energetic dispersal of fuel particles into the coolant (Reference Q12-3). The current pressurized water reactor regulatory acceptance criterion of 280 cal/gm is based primarily on experiments such as SPERT that were conducted by the Atomic Energy Commission. More recent REA experiments conducted at the Cabri facility in France, among others, suggest that a lower enthalpy limit may be appropriate, particularly for high 13
burnup irradiated fuel. The Electric Power Research Institute (EPRI) has used the more recent experimental data, coupled with cladding failure predictions using the Critical Strain Energy Density (CSED) approach, to develop proposed REA enthalpy limits as a function of burnup. The work is documented in EPRI's Topical Report on Reactivity Initiated Accident: Bases for RIA Fuel and Core Coolability Criteria" (Reference Q12-4),
which as been submitted to the Nuclear Regulatory Commission (NRC) and is currently under review.
Four MOX fuel rods have been tested under simulated REA conditions as part of the Cabri test program. Of those tests, three experienced no cladding failure with peak enthalpies of 138, 203, and 90 cal/gm. However, the Rep Na-7 test saw a cladding failure with fuel dispersal at an enthalpy of 120 cal/gm. The Rep Na-7 rod had a burnup of 55 GWd/MThm and a cladding oxidation layer of 50 microns (Reference Q 12-4, Table 2-1).
Based on the results of that test, it has been postulated that differences in fuel pellet microstructure between MOX and LEU fuel may make MOX fuel more susceptible to disruptive cladding failure at lower fuel pellet enthalpy values.
Accordingly, for the MOX fuel lead assemblies, Duke proposes to use a radial average peak fuel enthalpy limit that is substantially more conservative than the current NUREG-0800 acceptance criterion for LEU fuel. Duke proposes to use a value of 100 cal/gm at all burnups as the acceptance criterion for MOX fuel rods experiencing a power excursion from hot zero power (HZP). This criterion is considered to be appropriate and conservative, for the reasons provided below.
- 1. The value is significantly lower than enthalpies at which disruptive failure has been experienced in any MOX fuel REA tests.
- 2. The value is significantly lower than the Fuel Rod Failure Threshold curve for LEU fuel as proposed by EPRI (Reference Q12-4, Figure S-i).
- 3. MOX fuel rods will be clad in M5Tm. Fuel rod corrosion is considered to be a contributing factor to cladding failure under REA conditions. M5'm has demonstrated extremely low corrosion relative to Zircaloy-4, the cladding material that was used in all MOX fuel REA tests (see Figure 6.1 of Reference Q12-5).
- 5. Applying the criterion only to accidents from HZP excludes accidents initiating from hot full power with a high initial enthalpy (reflective of full power) but no rapid energy deposition in the fuel pellet.
Duke will use the SIMULATE-3K MOX computer code to perform three-dimensional reactor kinetics calculations of licensing basis REAs for all cores containing MOX fuel lead assemblies. Duke will verify that the peak enthalpy in all MOX fuel lead assembly rods remains below the 100 cal/gm acceptance criterion during postulated REAs.
SIMULATE-3K MOX, described in Section 2.4 of Reference Q12-6, is an extension of SIMULATE-3K. Application of SIMULATE-3K for REAs at Catawba has been reviewed and approved by the NRC (Reference Q 12-7) for cores containing LEU fuel.
Analyses of representative cores containing MOX fuel lead assemblies are summarized in 14
Section 3.7.2.4 of Reference Q12-8 and further detail will be provided in the response to Reactor Systems RAI Question 33.
The above criteria are conservative provisional criteria for the MOX fuel lead assembly program. To support the batch use of MOX fuel, Duke intends to propose alternative REA acceptance criteria. Duke plans to document the batch use MOX fuel REA acceptance criteria and REA analytical methodology in a MOX fuel safety analysis topical report and submit the report to the NRC for review in 2004.
References Q12-1. DPC-NE-300 I-PA, MultidimensionalReactor TransientsandSafety Analysis Physics ParametersMethodology, Duke Power Company, December 2000.
Q12-2. NUREG-0800, U. S. Nuclear Regulatory Commission StandardReview Plan, Revision 2, July 1981.
Q12-3. Meyer, R. O., McCardell, R. K., Chung, H. M. Diamond, D. J. and Scott, H. H.,
A Regulatory Assessment of Test Datafor Reactivity-InsertionAccidents, Nuclear Safety. Volume 37, No. 4, October-December 1996.
Q12-4. EPRI Technical Report 1002865, Topical Report on Reactivity Initiated Accident: Bases for RIA Fuel andCore CoolabilityCriteria,June 2002 (currently under NRC review).
Q12-5. BAW-10238(P), Revision 1, MOXFuel Design Report, Framatome ANP, May 2003 (currently under NRC review).
Q 12-6. DPC-NE- 1005P, Duke PowerNuclear Design Methodology Using CASM04/SIMUL4TE-3 MOX, August 2001 (currently under NRC review).
Q12-7. DPC-NE-2009-P-A, Revision 2, Duke Power Company Westinghouse Fuel TransitionReport, December 2002.
Q12-8. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.
Radiological Dose Criteria General radiological criteria are provided in 10CFR 20, 10CFR 50 Appendix A, 10CFR 50.67 and IOCFR 100. These are not published as uranium specific criteria, but have been consistently applied to reactor applications by the nuclear industry. Some of these regulations also apply to other applications, such as nuclear medicine. The applicable acceptance criteria in 10CFR are determined by the purpose or scenario for which the consequences must be calculated, rather than by the source term or specific isotopes involved.
The purpose of modeling the event and projecting consequences is to protect the health and safety of the public. To that end, there must be a standard for comparison to draw a definitive conclusion as to the impact upon the public. In order to compare the biological effects from the various isotopes which are produced in nuclear applications and 15
industries, the concept of dose equivalent (or committed dose equivalent) was adopted.
Usually expressed in Rems or Sieverts, these units provide a comparison of biological effects by accounting for the energy deposition and the relative biological effectiveness from radiation emitted by isotopes.
Since dose is a measure of the cumulative biological effect of the emitted particles and rays regardless of the isotope of their origin, there is no need to specify specific dose acceptance criteria for a reactor using MOX fuel. Furthermore, the criteria which are currently in regulations for the protection of the health and safety of the public and control room operators can be applied for the same purpose and application that they currently are being applied within a plant's licensing basis. The dose acceptance criteria in 10 CFR can be applied in the same manner as applied for LEU fuel. Standard Review Plan guidance can continue to be applied in accordance with a plant's licensing basis as it has been for LEU fuel. The specific regulatory dose criteria used to analyze MOX fuel events are summarized in Table Q12-3.
16
Table Q12-1 Applicability of 10CFR 50.46 Criteria to MOX Fuel Lead Assemblies 504 (b) 81,,,S, Apliabliy to MOX Fuel 0LeadAssemblie Bw _
This criterion concerns the performance of the fuel pin cladding material during LOCA and is, therefore, primarily related to cladding properties. The MOX lead assembly fuel rods will be constructed using Framatome ANP's M5'cladding.
The 2200 OF criterion has been approved by the NRC as applicable to M5Tm cladding in granting the licensing of replacement fuel for several light water Peak Clad reactors over the last few years. The basis for approval is experimental evidence Temperature that M57 behavior during LOCA conditions is equivalent to or superior to Zircaloy
< 2200 OF and is documented in BAW-1 0227P-A, 'Evaluation of Advanced Cladding and Structural Material (M5w) in PWR Reactor Fuel,' February 2000."
This temperature criterion has no dependence on the fuel pellet design or makeup and is equally applicable for use with either U0 2 or MOX fuel pellets.
This criterion is fully applicable to the MOX fuel lead assemblies.
This criterion concerns the performance of the fuel pin cladding material during LOCA and is, therefore, primarily related to cladding properties. The MOX lead assembly fuel rods will be constructed using Framatome ANP's M 5 claddin.
The 17 percent criterion has been approved by the NRC as applicable to M5 cladding in granting the licensing of replacement fuel for several light water reactors over the last few years. The basis for approval is experimental evidence 17% Local that M5 behavior during LOCA conditions is equivalent to or superior to Zircaloy don and is documented in BAW-10227P-A, 'Evaluation of Advanced Cladding and Structural Material (M5Tm) in PWR Reactor Fuel," February 2000."
The oxidation limit criterion controls the amount of hydrogen available to develop zirconium hydrides which increase the brittleness of the cladding in the post-accident environment. The criterion is not affected by the type of fuel pellet.
This criterion is fully applicable to the MOX fuel lead assemblies.
This criterion assures acceptable conditions within the reactor building and is
%Co unrelated to the core fuel and cladding so long as the hydrogen produced per i re- percent cladding reacted is unchanged. Because the reaction for both M5M and Oxida Zircaloy is between zirconium and oxygen, the hydrogen produced per reaction Oxldaion percent is the same for both materials. The criterion Is unaffected by the use of M5T cladding and is fully applicable to the MOX fuel lead assemblies.
Core This criterion controls the geometry of the core following a LOCA. As a criterion, it Amenable to achieves its purpose regardless of the cladding material or the fuel pellet makeup.
Cooling It is fully applicable to the MOX fuel lead assemblies.
17
Long-term This criterion controls the availability of long-term cooling systems and core Core conditions. As a criterion, it achieves its purpose regardless of the cladding Core material or the fuel pellet makeup. It is fully applicable to the MOX fuel lead assemblies.
18
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Transientl ccide- .:L5.~--Lx.F25: Accetnce Criteda- .-
- --description . .. ;4 "'a':
.2.1.3 LOCA Mass and Energy
- Containment design margin is maintained.
elease and Containment
- Environmental qualification of the safety related equipment inside 3ressure/Temperature Response containment Is not compromised.
5.2.1.4 Secondary System Pipe
- Containment design margin is maintained.
Ruptures and Containment
- Environmental qualification of the safety related equipment inside ressure/Temperature Response containment is not compromised.
15.1.1 Feedwater System alfunctions that Result In a
- Bounded by excessive increase In secondary steam flow analysis eduction in Feedwater in Section 15.1.2 and same criteria apply.
Temperature
- Peak RCS pressure remains below 110% of the design limit 5.1.2 Feedwater System (<2750 psia) alfunction Causing an Increase
- Fuel cladding Integrity shall be maintained by ensuring that the n Feedwater Flow calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110% of the design limit 15.1.3 Excessive Increase in (<2750 psia) econdary Steam Flow a Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110% of the design limit 15.1.4 Inadvertent Opening of a (<2750 psia)
Steam Generator Relief or Safety
- Fuel cladding integrity shall be maintained by ensuring that the Valve calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110% of the design limit
(<2750 psia)
- The potential for core damage Is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95195 DNBR 5.1.5 Steam System Piping limit based on an acceptable DNBR correlation. If the DNBR falls Failrem ibelow these values, fuel failure must be assumed for all rods that railuredo not meet these criteria. Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain In place and intact with no loss of core cooling capability.
- Offsite doses calculated shall not exceed the guidelines of 10CFRI00.
19
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies I;
Transient! cldent. .Acceptance Citeri-15.2.1 Steam Pressure Regulator
- Not applicable, there are no pressure regulators in the McGuire or Malfunction or Failure That Results Catawba plants whose failure or malfunction could cause a steam n Decreasing Steam Flow flow transient.
15.2.2 Loss of External Load
- Bounded by turbine trip analysis in Section 15.2.3 and same criteria apply.
- Peak RCS pressure remains below 110% of the design limit
(<2750 psia) 5.2.3 Turbine Trip
- Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
15.2.4 Inadvertent Closure of
- Bounded by turbine trip analysis in Section 15.2.3 and same Main Steam Isolation Valves criteria apply.
15.2.5 Loss of Condenser
- Bounded by turbine trip analysis in Section 15.2.3 and same acum and Other Events criteria apply.
Causing a Turbine Trip
- Peak RCS pressure remains below 110% of the design limit 15.2.6 Loss of Non-Emergency (<2750 psia)
C Power to the Station Auxiliaries
- Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110% of the design limit 15.2.7 Loss of Normal Feedwater (<2750 psia) iow
- Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110 % of the design limit
(<2750 psia) for low probability events.
15.2.8 Feedwater System Pipe
- Fuel cladding Integrity shall be maintained by ensuring that the Break calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- No hot leg boiling occurs.
- Peak RCS pressure remains below 110% of the design limit 15.3.1 Partial Loss of Forced (<2750 psla) 1eactor Coolant Flow
- Fuel cladding Integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
20
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Transient/AcdentAcceptance Criter a Descnption
- Peak RCS pressure remains below 110% of the design limit 15.3.2 Complete Loss of Forced (<2750 psia) eactor Coolant Flow
- Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Peak RCS pressure remains below 110% of the design limit
(<2750 psia) 15.3.3 Reactor Coolant Pump
- Any fuel damage calculated to occur must be of sufficiently limited haft Seizure (Locked Rotor) extent that the core will remain in place and intact with no loss of core cooling capability.
- Any activity release must be such that the calculated doses at the site boundary are a small fraction of the IOCFR100 guidelines.
15.3.4 Reactor Coolant Pump
- Bounded by reactor coolant pump shaft seizure analysis in Shaft Break Section 15.3.3 and same criteria apply.
- Peak RCS pressure remains below 110% of the design limit 15.4.1 Uncontrolled Rod Cluster (<2750 psia) ntrol Assembly Bank
- Fuel cladding Integrity shall be maintained by ensuring that the ithdrawal From a Subcritical or calculated DNB ratio remains above the 95/95 DNBR limit based ow Power Startup Condition. on an acceptable DNBR correlation.
- Fuel centerline temperatures do not exceed the melting point
- Peak RCS pressure remains below 110% of the design limit 15.4.2 Uncontrolled Rod Cluster (<2750 psia)
Control Assembly Bank
- Fuel cladding integrity shall be maintained by ensuring that the Withdrawal at Power calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
- Fuel centerline temperatures do not exceed the melting point.
- Peak RCS pressure remains below 110%/6 of the design limit 15.4.3 Rod Cluster Control (<2750 psia)
Assembly Misoperation (System cFuel cladding integrity shall be maintained by ensuring that the Malfnctin orOpeatorErro) -calculated DNB ratio remains above the 95/95 DNBR limit based alftDrop on an acceptable DNBR correlation.
- Fuel centerline temperatures do not exceed the melting point
- Peak RCS pressure remains below 110% of the design limit
(<2750 psia) 15.4.3 Rod Cluster Control
- Fuel cladding integrity shall be maintained by ensuring that the Assembly Misoperation (System calculated DNB ratio remains above the 95195 DNBR limit based Malfunction or Operator Error) - on an acceptable DNBR correlation.
Single Rod Withdrawal
- Fuel centerline temperatures do not exceed the melting point
- Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10CFR100 guidelines.
21
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Tansient/Accident Accep nce reria y Description ________________-___:__________-__:________-___L__
- Peak RCS pressure remains below 110% of the design limit 15.4.4 Startup of an Inactive (<2750 psia)
Reactor Coolant Pump at an
- Fuel cladding integrity shall be maintained by ensuring that the noorrect Temperature calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
15.4.6 Chemical and Volume a Peak RCS pressure remains below 110% of the design limit Control System Malfunction that (<2750 psia)
Results in a Decrease in Boron
- Fuel cladding integrity shall be maintained by ensuring that the Concentration in the Reactor calculated DNB ratio remains above the 95/95 DNBR limit based Coolant on an acceptable DNBR correlation.
5.4.7 Inadvertent Loading and
- Any activity release must be such that the calculated doses at the ration of a Fuel Assembly in site boundary are a small fraction of the 10CFR1 00 guidelines.
an Improper Position
- Peak RCS pressure remains below 120% of design for very low probability events (< 3000 psia).
- Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limit based 15.4.8 Spectrum of Rod Cluster on an acceptable DNBR correlation.
Control Assembly Ejection
- Any fuel damage calculated to occur must be of sufficiently limited Accidents extent that the core will remain in place and intact with no loss of core cooling capability.
- The fission product release to the environment is well within the established dose acceptance criteria of 10CFR100.
- See provisional caVgm acceptance criteria attached.
- Peak RCS pressure remains below 110% of the design limit 15.5.1 Inadvertent Operation of (<2750 psia)
- Fuel cladding integrity shall be maintained by ensuring that the During Power Operation calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
15.5t2 Chemical and Vonume
- Bounded by inadvertent operation of emergency core cooling nola ste ealu colamn That system during power operation analysis in Section 15.5.1 and nventoryesame criteria apply.
- Peak RCS pressure remains below 110% of the design limit 15.6.1 Inadvertent Opening of a (<2750 psia)
Pressurizer Safety or Relief Valve
- Fuel cladding integrity shall be maintained by ensuring that the calculated DNB ratio remains above the 95/95 DNBR limidt based on an acceptable DNBR correlation.
15.6.2 Break In Instrument Line or Oher Lines From Reactor Coolant
- Any activity release must be such that the calculated doses at the Pressure Boundary That Penetrate site boundary are a small fraction of the 10CFR100 guidelines.
Containment 22
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies TranseitlAccident A e a iteria
^ >Description s if-; ecetne w.E^f+ -Ats
- Fuel cladding integrity shall be maintained by ensuring that the 15.6.3 Steam Generator Tube calculated DNB ratio remains above the 95/95 DNBR limit based Failure on an acceptable DNBR correlation.
- Any activity release must be such that the calculated doses at the site boundary are a small fraction of the IOCFR100 guidelines.
23
Table Q12-3 Regulatory Dose Criteria For Accidents with MOX Fuel Lead Assemblies Of ; Reference SourceTem
Reference:
300R R Th) d RGs1.195 RG 1.183 LOCA25 Rem WB SRP115.6.5 App. A 10CFR50.67 RG 1.195 Steam Generator Tube Rupture 300 Rem Thyroid 10CFRO1O.195 25 Rem TEDS RG 1.183 with fuel failure or pre-incident iodine spike 25 Rem WB SRP 15.6.3 25_emTEER_118 SRP 15.195 Steam Generator Tube Rupture 30 Rem Thyroid 1CFRO1O.151/ 2.5 Rem TEDE RG 1.183 with concurrent Iodine spike 2.5 Rem WB SRP 15.6.3 10CFR50.67 SRP 15.195 Main Steam Line Break 300 Rem Thyroid RG 1.195, 25 Rem TEDE RG 1.183 with fuel failure or pre-incident Iodine spike 25 Rem WB SRP 15.1.5 App. A 210CFR50.67 SRG 15.195 Ap Main Steam Line Break 30 Rem Thyroid RG 1.195 . 2.5 Rem TEDE RO 1.183 with concurrent Iodine spike 2.5 Rem WB SRP 15.1.5 App. A 2_5_Rem___ _ 1_183 RG SRP1.19583 Locked Rotor Accident 30 Rem Thyroid 2.5 Rem WB 15.3.32.ReTDEG118 Rod EjectinRdEetoAcdet75 Accident 6.3 Rem Rem Thyroid WB2 RGI1.195 SRP 15.4.8 App A 6.3 Rem TEDE RG 1.183 FulHnln ciet75 Rem Thyroid RG 1.195 Fuel Handling Accident 6.3 Rem WB2 SRP 15.7.4 6.3 Rem TEDE RG 1.183 Ro ODses Conro 50 Rem Tyold I3 RG 1.195.183 All 5 Rem WB 10CFRIO/ 5 RemITEDE RG 1.183 50 Rem 1 skin 50Resin Appendix lGDC 19ll AN 5ORem5TEDS WB= Whole body, RG--Regulatory Guide, SRP= Standard Review Plan 2 Where a conflict exists between SRP and RG 1.195 on the whole body dose limit for a particular accident, the more current guidance is shown.
3 RG 1.195 specifically states that this criterion may be used in lieu of the one in the SRP.
24
- 13. To allow the NRC staff to perform confirmatory analysis, please provide both the McGuire and Catawba loss-of-coolant accident (LOCA) input decks for the low enriched uranium (LEU) as well as the MOX fuel rods. Provide the decks in an electronic format, including nodalization diagrams.
Response (Previously submitted October 3, 2003)
The accompanying compact disc includes two RELAP5/MOD2-B&W input decks in UNIX format as follows:
r5moxnrc.in - Input deck for MOX fuel pins, power peaked at 10.3 ft.
r5uo2nrc.in - Input deck for LEU fuel pins, power peaked at 10.3 ft.
These are blowdown input decks used in the deterministic evaluations of MOX and LEU fuel pins reported in the license amendment request. The deterministic MOX fuel calculations comprise the licensing basis for the MOX fuel lead assemblies.
Deterministic LEU fuel calculations were included to address the relative LOCA performance between MOX and LEU fuel.
Figures Q13-1 and Q13-2 are node diagrams for the decks. Figure Q13-1 shows the loop node arrangement while Figure Q13-2 shows the reactor vessel node arrangement.
Figure 3-5 of Attachment 3 to Reference Q 13-1 provides some additional detail specific to the core region.
RELAP5/MOD2-B&W is a derivative of the INEL code RELAP5/MOD2. Many changes were made to the INEL code to create the approved Framatome ANP deterministic LOCA code. Because the input for these changes may not be recognizable by other versions of RELAP5, the following list of related input card images is provided to assist the NRC staff.
Card 190: EM Choking Model Specification Card (Activates Framatome ANP specific choked flow break modeling.)
Card 192: EM Critical Flow Transition Data (Activates Framatome ANP specific critical flow break modeling.)
Card 195: Interface Heat Transfer Weighting (Activates Framatome ANP specific interface heat transfer weighting.)
Cards 10000020-10000029: Heat Structure Cards (Activate Framatome ANP specific filtered flow model - 10CFR50.46 Appendix K requirement.)
Cards 10000S80-10000S99: Reflood Grid and Wall Heat Transfer Factor Data (Activate Framatome ANP specific grid model for droplet breakup and convective heat transfer due to grids.)
Cards lCCCG801-ICCCG899: Left Boundary Heat Structure Cards Cards lCCCG901-ICCCG999: Right Boundary Heat Structure Cards 25
(Activate the Framatome ANP specific EM heat transfer package.)
Cards 19997000-19999999: EM Pin Model Specification (Activate Framatome ANP specific EM core package providing for dynamic fuel-clad gap conductance and fuel rod swell and rupture. Also provide the M5S' cladding properties.)
Reference Q13-1. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.
26
Figure Q13-1 Loop Noding Diagram 3 e Is 6455 624 63 =
625-1 633.2 .
625-2 633-1 s I
or Vesge CMW Triple Loop Single loop 27
Figure Q13-2 Core Noding Diagram I
364 ] Upper Head I
360 Upper Plenum 358 300 Outlet Annulus CV 170 II CV 100 CV 280 CV 200 1352-5 1 445-1 345.1 445 345 444 344 443 343 441 341 350-3 346-3 304 439 339 438 338 437 337 436 336 435 335 434 334 C r
LJ & sa 433 432 333 332 a 331 I 346-2 F7 350-2 b 431 U 430 330 t IL II 428 328 427 327 426 425 325 306 324 423 323 422 322 350-1 421 321 346-1 320 419 319 417 317 1 314-1 315-1 314 308 Lower Plenum 312 1 310 l 28
- 14. Provide the reference to the best estimate LOCA model noted in section 3.7.1.7.
Response (Previously submitted October 3, 2003)
Based on RAI Questions 14, 15, and 16 it appears that some clarification is needed with respect to the LOCA analysis performed for the MOX fuel lead assemblies and how this analysis is used to support the lead assembly cores. In summary, the licensing basis for the resident Westinghouse RFA fuel remains the best estimate large break LOCA analysis performed by Westinghouse. Framatome ANP Appendix K analyses demonstrate that changing the fuel pellet material to MOX has no significant impact on peak cladding temperature following a large break LOCA. Framatome ANP Appendix K analyses provide peaking limits that ensure the peak cladding temperature for MOX fuel rods following a large break LOCA remain within the regulatory limit. The following discussion provides a further description of the analysis performed for the resident fuel assemblies as well as the MOX fuel lead assemblies.
Resident Fuel The resident fuel in MOX fuel lead assembly cores will be the robust fuel assembly (RFA) design that is supplied by Westinghouse. The large break LOCA analysis that supports this fuel design is the Westinghouse best estimate method described in Reference Q14-1. The analysis is based on the WCOBRA/TRAC method and includes detailed treatment of the uncertainties associated with the computer models and the inputs related with plant operation. As part of the analysis, Westinghouse performed sensitivity studies to address transition or mixed core effects. This was necessary because the RFA fuel was initially introduced into cores containing Framatome ANP Mark-BW design fuel. The conclusion of the mixed core sensitivities was that the presence of the Mark-BW fuel assemblies had an insignificant impact on the calculated results. Westinghouse also performed small break LOCA calculations for McGuire and Catawba using the NOTRUMP methodology as described in Reference Q14-2. A mixed core penalty of 100F was assessed and applied to the small break LOCA results to accommodate the presence of the Mark-BW fuel assemblies. Given that the MOX fuel lead assemblies are more similar hydraulically to the RFA fuel than the Mark-BW design fuel, the mixed core penalty developed for the Mark-BW fuel assemblies bounds the MOX fuel lead assemblies. Therefore, the Westinghouse LOCA analyses for the resident RFA fuel remain valid in the presence of four MOX fuel lead assemblies.
MOX Fuel Lead Assemblies To address the MOX fuel lead assemblies, Framatome ANP performed deterministic large break LOCA calculations consistent with the requirements of 10 CFR 50 Appendix K. In order to model accurately the effect of changing the fuel pellet material to MOX, Framatome ANP made modifications to their deterministic large break LOCA method as described in Reference Q 14-3. These modifications are described in Section 3.7.1.2 of Attachment 3 to Reference Q14-4. Next, Framatome ANP performed large break LOCA calculations for a MOX fuel lead assembly as well as a Framatome ANP LEU fuel assembly, with both analyses assuming the hydraulic characteristics of the Advanced Mark-BW fuel assembly design. This sensitivity study was performed to assess the impact of the change in fuel rod parameters (MOX vs. LEU) on the calculated results. As discussed in Section 3.7.1.3 of Attachment 3 to Reference Q14-4, this sensitivity study showed that there is essentially no difference between the LOCA results for the MOX 29
fuel and the LEU fuel (APCT of 370 F). The Framatome ANP MOX fuel lead assembly results were also compared to the Westinghouse best estimate results to illustrate the similarity of the results. Given the differences in the two analytical methods, a direct comparison of the results is not completely valid. However, the comparison illustrates that the MOX fuel lead assembly with the lower peaking assumptions yields lower peak cladding temperature results (APCT of-387F).
Following submittal of the MOX fuel license amendment request, Framatome ANP completed additional cases to investigate the impact of steam generator type, time in life, and axial power shape. Two different steam generator designs were examined:
Westinghouse Model D5 steam generators (Catawba Unit 2), with a 10% tube plugging assumption; and BWI steam generators (Catawba Unit 1), with 5% tube plugging. The study concluded that the Model D5 steam generators with the 10% tube plugging assumption are limiting with respect to the Framatome ANP deterministic large break LOCA analysis. The D5 case provided the base case input for the other sensitivities cases.
Framatome ANP performed time in life sensitivities to assess the large break LOCA results as the stored energy in the fuel rod varies with cycle burnup. At burnups greater than 30 GWd/MThm, a Kau factor is applied to limit the PCT for these cases. The KBu factor reduces the FQ (total peaking factor) as well as the FAh (enthalpy rise factor or radial peaking factor).
Furthermore, using the limiting burnup case which uses a KBU of 1.0, i.e., the 30 GWd/MThm case, Framatome ANP evaluated power peaks at different elevations. The purpose of these sensitivities was to establish LOCA limits as a function of core height.
At elevations above the 8 foot elevation a Kz factor was applied. The Kz factor reduces the FQ as well as the axial peaking factor (Fz).
A summary of the sensitivity cases is provided in Table Q14-1. The resulting LOCA peaking requirements for the MOX fuel lead assemblies are shown in Figure Q14-1.
These peaking requirements will assure that the MOX fuel will comply with the regulatory limits for LOCA as provided in the response to Reactor System RAI Question 12.
MOX Fuel Lead Assembly Licensing Basis The licensing of the MOX lead assemblies will be based on analysis to determine the relative accident performance between the MOX and resident LEU assemblies because of the different fission source materials. As presented in the license amendment request, large break LOCA calculations, using the Framatome ANP deterministic LOCA evaluation model, have been performed for both LEU and MOX assemblies. The LEU calculations applied the evaluation model as approved by NRC. The MOX calculations applied the evaluation model with specified alterations, described in the LAR, necessary to simulate MOX fuel. The comparison of these two calculations demonstrated the expected result: that there is essentially no difference in the large break LOCA performance between fuel, of comparable design, using MOX pellets and fuel using LEU pellets. An evaluation of the small break LOCA, provided in the LAR, also determined that there would be no differences in the calculated results between the MOX and LEU fuel assemblies. Therefore, the assessment of the Catawba LOCA performance for the 30
cores with four MOX lead assemblies is that LOCA performance is not altered. This result, in combination with a reduction in the allowed peaking factor for the MOX fuel pins, provides the licensing basis for the MOX fuel lead assemblies assuring that all of the criteria of 10CFR50.46 are met.
References Q14-1. WCAP-12945P-A, Volume 1 Revision 2 and Volumes 2-5 Revision 1, Code QualificationDocumentforBest-Estimate Loss of CoolantAnalysis, March 1998.
Q14-2. WCAP- 100564P-A, Westinghouse Small BreakECCS Evaluation Model using the NOTRUMP Code, August 1985.
Q14-3. BAW-10168P-A, Revision 3, RSGLOCA -BWNTLoss-of-Coolant Accident Evaluation Modelfor Recirculating Steam GeneratorPlants, December 1996.
Q14-4. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.
Table Q14-1 Summary of MOX Fuel Lead Assembly Large Break LOCA Sensitivity Cases Model D5 SGs with 10% Tube Plugging TIL. Elevation. -
-. (GWd/whri) = Kz_CT Fz Eq ('F)
BOL 6.8556 1.0 1.0 1.6 1.500 2.4 1919.2 20 6.8556 1.0 1.0 1.6 1.500 2.4 1943.6 30 6.8556 1.0 1.0 1.6 1.500 2.4 1948.8 50 6.8556 0.867 1.0 1.387 1.500 2.08 1824.4 60 6.8556 0.8 1.0 1.280 1.500 1.92 1787.6 30 4.7001 1.0 1.0 1.6 1.500 2.4 1815.0 30 8.5656 1.0 0.993 1.6 1.490 2.383 1964.0 30 10.2756 1.0 0.972 1.6 1A58 2.332 2019.5 31
Figure Q14-1 MOX Fuel Lead Assembly Total Core Peaking Factor 2.5 2A 2.3 2.1 alft
-*-8.5656ft 1~-0.2756 ft ft 1i-2 1.7 0 10 20 30 40 50 60 Bumup (GWd/MThm) 32
- 15. Provide the uncertainty analysis that was performed for the LEU and MOX LTA demonstrating that the 95/95 peak cladding temperature has been calculated for the core.
The response is expected to include a complete discussion of the statistical methodology used.
Response (Previously submitted October 3, 2003)
The MOX fuel and LEU fuel LOCA analyses that support the use of the MOX fuel lead assemblies are deterministic calculations and therefore no uncertainty analysis was performed. See the response to Reactor Systems RAI Question 14 for additional explanation.
Provide the Phenomena Identification and Ranking Table for the LOCA analyses performed with the best estimate model and reference the best estimate model used for the analysis.
Response (Previously submitted October 3, 2003)
The Phenomena Identification and Ranking Table (PIRT) used in the Westinghouse best-estimate LBLOCA analysis is contained in Reference Q16-1. Since this method was not used to directly support the MOX fuel lead assemblies, this PIRT is not applicable to the MOX fuel lead assembly analysis. See the response to Reactor Systems RAI Question 14 for additional explanation.
Reference Q16-1. WCAP-12945P-A, Volume 1 Revision 2 and Volumes 2-5 Revision 1, Code QualificationDocumentfor Best-Estimate Loss of CoolantAnalysis, March 1998.
- 17. Provide the experimental data base used to assess the biases and to determine the uncertainties in the fuel rod behavior for the MOX LTA.
Response (Previously submitted October 3, 2003)
The database is provided in Chapter 3 of the COPERNIC topical report (Reference Q1 7-1). Additionally, at the NRC's request, several MOX fuel rods from the Halden experiments were analyzed with COPERNIC to end-of-life burnups in the range of 50 to 64 GWd/MThn.
Reference Q17-1. BAW-10231P Revision 2, COPERNICFuel RodDesign Computer Code, July 2000.
- 18. In sub-section 3.7.1.1.1, nothing is mentioned about the MOX/LEU interface behavior.
Provide a qualitative and quantitative discussion regarding the neutron flux behavior at the interface of the MOX and LEU fuel assemblies.
Response (Previously submitted October 3, 2003)
Duke used the CASMO-4 computer code to model pin cell neutron flux and power at the intersection of four quarter-assembly lattices. These "colorsets" provide detailed two dimensional neutronic calculations that account for interface effects between dissimilar fuel assemblies. MOX fuel assemblies and LEU fuel assemblies of equivalent lifetime 33
- 21. How does the lower fuel conductivity of the MOX fuel impact the maximum pellet centerline temperature during a LOCA as compared to LEU fuel? Please provide a qualitative and quantitative discussion of the differences.
Response
There is only a slight difference in the fuel pellet conductivity between MOX fuel of the lead assembly design and plutonium concentration and comparable LEU fuel. Figure Q21-1 compares the thermal conductivity for MOX fuel pellets of the lead assembly design to comparable LEU fuel pellets for both un-irradiated fuel and fuel irradiated to 40 GWd/MThm. The thermal conductivity values shown in Figure Q21-1 are from the fuel performance code COPERNIC (Reference Q21-1). COPERNIC has been approved by NRC for use with LEU fuel and is under review for MOX fuel applications. Although thermal conductivity values in Figure Q21-1 change with burnup for both MOX fuel and LEU fuel, the offset, approximately two percent, is constant.
The analyses presented in Section 3.7.1 of Attachment 3 to the license amendment request (Reference Q21-2) directly compare the effect of the MOX to LEU offset in conductivity in conjunction with the other differences in the fuel pin designs. Figures Q21-2 and Q21-3 provide a fuel pin temperature profile comparison between MOX and LEU fuel pellets at the accident initial conditions and at the approximate time of peak cladding temperature. As expected, there is little difference in the temperature distributions between the two fuel types. Figure Q21-4 provides the evolution of the centerline fuel temperatures with time for the MOX and LEU fuel at the location of peak cladding temperature. The two fuel temperatures differ slightly during the course of the transient.
The variation is attributed to fuel pellet thermal conductivity and to other differences in the fuel pin design. As an example, the LEU fuel pin has a higher pre-fill pressure than the MOX pin. The higher pressure increases the hoop stress resulting in a slightly lower calculated rupture temperature and earlier calculated rupture time. Combined with all of the models interacting to determine the cladding and pellet temperatures the LEU fuel centerline temperature is 40'F cooler at the time of peak cladding temperature.
The difference in thermal conductivity between MOX fuel of the lead assembly design and comparable LEU fuel is small. The effect of this difference on LOCA calculational results is nil and not distinguishable from the effects of normal fuel design variations.
References Q21-1. BAW-1 0231 P Revision 2, COPERNICFuel Rod Design Computer Code, July 2000.
Q21-2. Tuckman, M. S., February 27, 2003, Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.
41
Figure Q21-1 Thermal Conductivity Comparison for MOX and LIEU Fuel (Fuel porosity of 0.0479)
QOPr..nd a BLEU@O0GWdfmntU -
I a a a ILU@~400GWd/nV a a a a a '4 vA/9MOX@O 0GWdlmnthm a a a I
7.OE-04 a aKa -4 wt%/MOX@~40 a a i6.OE-04 I IS.OE-04 U
a a a tr aha a
-a-a-a-a-a-aa a a I 40OE.04 3.OE-04 2.OE-04 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Fuel1Temperammr (F) 42
Figure Q21-2 MOX and LEU Fuel Pin Temperature Profile Comparison at Loss of Coolant Accident Initiation 2500 -----------
2000 __--_--_--_ t ---
[---
2500 30 0-t -----
X--
-- - -- - H-100ow-- LEU - -- -- -- -- -- -- -- - - - ----- -< -- - t - -- -- -r- --
-e- 4 wel/ MOX l- -
MOO. .. I.I 0 1 2 3 4 5 6 7 8 9 10 Radial Mesh Point (#)
43
Figure Q21-3 MOX and LEU Fuel Pin Temperature Profile Comparison at Time of Peak Cladding Temperature 3500 20001 - --- ----
1500 --------
I , I , I I 1000 I a LEU PCT time -125 sece , i
- 44wt*MOXPCrtime- 131we 500 I I I 0 1 2 3 4 5 6 7 8 9 10 Radial Mesh Point (#)
44
Figure Q21-4 MOX and LEU Fuel Centerline Temperature Comparison for Loss of Coolant Accident 3500 3000 -- - - - - - - - - - - - - - - - - - - - - - - - - - --r - - - - - - - --- - - - - - - - - - - - - - - - - - - - - - - -
p2500 2000 -d - -- -- - - - -_
1500 LEU (Fq=2.4) 4 wtff, MOX (Fq=2.4) 500 0 50 100 150 200 250 300 350 400 Time (seconds) 45
- 22. The first paragraph of section 3.7.1.1.2 states that "The result, including appropriate uncertainties, is that .." Please state the uncertainties that are being referred to in this section along with what is considered to be appropriate.
Response
References Q22-1 and Q22-2 are industry standard tools for calculating decay heat for low-enriched uranium (LEU) cores. Analysis of highly burned LEU fuel shows that it produces the majority of its energy from the fission of plutonium isotopes. Therefore, these standard tools are appropriate for calculating decay heat in cores containing MOX fuel and for determining the uncertainties to be applied.
The uncertainties included in the MOX fuel decay heat analysis include:
(1) ANSI/ANS-5 .1-1994 standard uncertainties for infinite irradiation by isotope, (2) ANSIIANS-5.1-1994 "ISO standard" for energy released from fission (the "Q" value),
(3) ANSI'ANS-5.1-1994 standard for absorption burnup correction factors, Gmax (t), and (4) actinide decay uncertainties.
Many of these values are a function of time after shutdown. Table Q22-1 shows the effect of time after shutdown on each of these uncertainties.
To obtain a reasonable statistical (95/95) tolerance/confidence factor to apply to the one sigma uncertainty, the Appendix K requirement and the standards were examined. As explained in the ANSI/ANS-5.1-1994 standard, the 1.2 uncertainty factor was based on work reported in the Bettis Technical Review by K. Shure. Shure's work stated that a relative uncertainty of 20% would bound all positive deviations in decay periods less than 107 seconds. The measured data indicate that the one sigma uncertainty is about 10%. Thus, there is a factor of two in the Appendix K requirements between the sigma and the bounding value. This implies that a tolerance/confidence factor of two is acceptable to use as a 95/95 percent level of confidence in the determination of conservative decay heat calculations. The MOX fuel decay heat model uses a tolerance/confidence factor of two applied to the uncertainties.
The 95/95 actinide decay heat fraction and the 95/95 fission product decay heat fraction are calculated and summed to produce the MOX fuel decay heat model. Comparing the results of the 95/95 MOX model with the standard Appendix K decay heat model for LEU fuel shows that the LEU model produces higher values of decay heat than MOX fuel. This is shown in Figure 3-3 of Attachment 3 of Reference Q22-3 for LOCA-typical decay times.
References Q22-1. American National Standard for Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1-1994, American Nuclear Society, 1994.
Q22-2. O.W. Hermann, R.M. Westfall, ORIGEN-S: Scale System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission ProductBuildup and Decay, andAssociatedRadiation Source Terms, NUREG/CR-0200, September 1998.
46
Q22-3. Tuckman, M.S., February 27, 2003, Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50.
47
Table Q22-1 Effect of Time after Shutdown on Decay Heat Uncertainty Factors Uncertainty Uncertainty Value Uncertainty Range (From Text Parameter (1.0 Second after (100 to I 0 Seconds Reference of Q22) Shutdown) after Shutdown)
(1) One sigma uncertainty for 235U 2.8% 1.7 - 2.8% ANSI IANS-5.1-1994, Page 14 fission product decay heat (1) One sigma uncertainty for 2"U 9.0% 3.8 - 9.0% ANSI /ANS-5.1-1994, Page 18 fission product decay heat (1) One sigma uncertainty for 239Pu 4.5% 3.6- 5.3% ANSI /ANS-5.1-1994, Page 16 fission product decay heat (1) One sigma uncertainty for 241Pu 5.4% 4.4- 10.0% ANSI /ANS-5.1-1994, Page 20 fission product decay heat (2) Q- sigma for 2U (MeV per Fission) +0.5 NA ANSI/ ANS-5.1-1994, Page 38 (2) Q- sigma for 234U (MeV per Fission) +1.0 NA ANSI/ ANS-5.1-1994, Page 38 (2) 0- sigma for 239Pu (MeV per +0.7 NA ANSI /ANS-5.1-1994, Page 38 F ission) __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
(2) 0- sigma for 24 1Pu (MeV per +0.7 NA ANSI/ANS-5.1-1994, Page 38 Fission) __ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _
(3) Gma (t) (Note1) 2% 2.0-18.1% ANSI/ANS-5.1-1994, Page26 23 9 U decay heat one sigma 10% NA Note 2 (4) uncertainty 239
() Np decay heat one sigma15NAoe2 (4) uncertainty% NA Note 2 (4) Decay heat for all other actinides 20% NA Note 2 one sigma uncertainty NA - Not applicable because there Is no apparent time dependence of this parameter.
Note 1: Gmax(t) Is the maximum correction relative to the nominal value of G(t).
Note 2: The actinide decay heat uncertainties are estimated based on the accuracy of ORIGEN-S and measured data.
48
- 24. Section 3.7.1.1.4 discusses the LOCA transient initialization and the changes made to accommodate using the COPERNIC code instead of the TAC03 code, including the adjustments made to some of the parameters. Provide additional information on the adjustments made, how the adjustments were developed and include any data used to develop the adjustment. Additionally, since these values are used in RELAP5 initialization, please show that throughout the fuel lifetime, the TAC03 and COPERNIC codes predict consistent values for the different fuel parameters used as input for the LOCA analysis.
Response
The discussion in Section 3.7.1.1.4 of Attachment 3 to Reference Q24-1 concerns alterations in the approach used to determine the fuel-to-clad gap conductance and in the values used for the initial fuel temperatures in the three core heat structures of the LOCA simulation. The approach to the fuel-to-clad gap conductance is described in detail in the response to Reactor Systems Question 25. The following discussion presents additional detail regarding the determination of the initial fuel temperatures for the core heat structures.
Because COPERNIC is NRC-approved for LOCA application to LEU fuel and includes modeling for MOX fuel properties, it was selected for the prediction of initial fuel temperatures for the MOX simulations and for the LEU comparison case. COPERNIC is an advanced fuel performance code relative to TAC03 and predictive consistency between COPERNIC and TAC03 should not be expected.
The Framatome ANP deterministic LOCA evaluation model, used to evaluate the MOX fuel lead assemblies, incorporates a two coolant channel, three heat structure core model to assure that the coolant and pin conditions for the hot spot are appropriate. The two coolant channels represent flow in the average core and flow in the hot fuel assembly respectively. The three heat structures represent the average core, the hot bundle, and the hot pin. Both the hot bundle and the hot pin couple thermal-hydraulically with the hot fuel assembly fluid channel. Figure 3-5 of Attachment 3 to Reference Q24-1 illustrates the arrangement. The NRC approved this core representation in, Reference Q24-2.
LOCA calculations include provision for appropriate uncertainties in both transient and initial conditions. One of those uncertainties is the initial fuel temperature or initial stored energy used in the core simulation. To determine the initial fuel temperatures, an NRC-approved fuel performance code, such as COPERNIC or TACO3, is run in accordance with the plant boundary conditions and core power distributions to be simulated. These codes produce best estimate predictions of the core temperature distributions that are transferred, after adding appropriate prediction uncertainties, to RELAP5/MOD2-B&W for the LOCA calculations. The uncertainties are determined from the benchmarks of the fuel performance codes and the make-up of the core region being modeled in RELAP5/MOD2-B&W.
For the hot pin, the LOCA calculation resolves a conservative representation of a single region of fuel pellets in a single rod. The appropriate level of uncertainty to add to the hot 50
pin initial temperature prediction is a temperature increment that gives a 95/95 confidence that the resultant temperature is not under predicted. For COPERNIC, the fuel performance code used for MOX simulations, this would comprise an addition of [ ]
to the prediction of the fuel temperatures along the entire hot pin. For a TAC03-based evaluation, 11.5 percent of the predicted fuel temperature would be added.
For the average core, the LOCA calculation resolves a representation of a large group of fuel pellets in many rods. The appropriate level of uncertainty to add to the initial temperature predictions includes the integration of individual pellet uncertainties over this entire group and a determination of the 95/95 confidence band for the entire group. With the size of the group involved, the aggregate uncertainty is near zero and it is appropriate to initialize this group, the average core, at the fuel performance code prediction without adjustment. With this selection, the COPERNIC [ ] the benchmark temperatures is conservatively ignored.
The more interesting initialization is that for the hot bundle representation. The purpose of the hot bundle is to provide the coolant conditions with which to cool the hot pin. As such, the hot bundle configuration is selected to represent the aggregate of the eight fuel pins immediately surrounding the hot pin. For TAC03, the appropriate 95/95 confidence level for the aggregate initial temperature or stored energy of a group of eight pins requires that the TAC03 prediction be increased by about 2.5 percent. The modeling approved by the NRC in Reference Q24-2 stipulated that the temperature prediction be increased by 3.0 percent to provide a small additional conservatism. The determination of the increase is dependent on the distribution of the uncertainty and bias for the fuel performance code. The TAC03 uncertainty distribution is a Gaussian or normal distribution and the difference in a temperature adjustment to achieve 95/95 confidence between a single member set and the set representing the eight fuel pins surrounding the hot pin is significant, 11.5 percent for the hot pin and about 2.5 percent for the surrounding pins. If the uncertainty distribution for COPERNIC is close to Gaussian, there will be little difference in the relative temperature adjustment. That is, the appropriate adjustment will be the same fraction of the 95/95 adjustment factor for both codes. In determining the uncertainty adjustment for COPERNIC applications, it was assumed that the COPERNIC uncertainty distribution was sufficiently close to Gaussian to employ this logic.
The justification of this argument only requires that the distribution of uncertainty for COPERNIC be reasonably normal and that the temperature adjustment providing a 95/95 confidence for a single member set be known. That the COPERNIC uncertainty is reasonably normal can be observed in a comparison of the TAC03 uncertainty distribution to a histogram of the COPERNIC benchmarks. This comparison is presented in Figure Q24-1 as normalized predicted minus measured data. By observation, the uncertainty distribution for COPERNIC, if correlated, would not differ markedly from that of TAC03 except for a slightly different bias. Thus, for the LOCA evaluation of the MOX lead assemblies, the COPERNIC prediction of the hot bundle initial temperature was increased by the ratio of hot bundle to hot pin adjustment for TAC03 times the hot pin adjustment for COPERNIC, 3.0/11.5 times [ ].
51
Reference Q24-1. Tuckman, M. S., February 27, 2003, Letter to U.S. Nuclear Regulatory Commission, Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50 Q24-2. Letter, U.S. Nuclear Regulatory Commission to Framatome ANP, Safety Evaluation of FramatomeTechnologies Topical Report BA W-10164P Revision 4, RELAP5MOD2-B&W, An Advanced Computer ProgramforLight Water Reactor LOCA andNon-LOCA TransientAnalysis, April 9, 2002.
Figure Q24-1 TACO Uncertainty Distribution Compared to COPERNIC Benchmark Histogram 52
- 25. Section 3.7.1.1.4 discusses RELAP5 initialization, stating that the core model will not be in steady state at transient initialization. Since a false declared steady state can lead to errors from an imbalance, please provide justification for why the RELAP5 model will not be in steady state at transient initiation and how steady state conditions for initialization are assured.
Response
The RELAP5/MOD2-B&W code includes a fuel pin model that represents the fuel rod in accordance with the requirements of 10CFRSO Appendix K. This model explicitly considers the fuel pellet, fuel-to-clad gap and clad-to-coolant heat transfer. It allows for specification of material conductivities for the pellet, gap, and cladding. The gap conductance term accounts for gaseous conductance, fuel pellet-to-cladding contact and radiation.
The initial fuel thermal conditions for LOCA are determined by an NRC-approved steady-state fuel performance code. For the analysis of the MOX lead assemblies, COPERNIC is used. The following input from COPERNIC is transferred to RELAP5/MOD2-B&W:
- Fuel rod temperatures after adjustment for uncertainties (29 axial and 10 radial nodes),
- Fuel pellet and cladding radial geometry,
- Fuel-to-cladding contact pressure,
- Initial internal fuel pin pressure,
- Fuel thermal conductivity, and
- Gas composition.
COPERNIC provides a best-estimate calculation of the initial fuel temperature distributions. To provide suitable inputs for RELAP5/MOD2-B&W, appropriate uncertainties are added to the predicted temperatures when they are transferred. This increase in temperature combined with the fact that COPERNIC and RELAP5/MOD2-B&W have slightly different gap models means that the steady-state initial fuel temperature predictions for the two codes will differ. Previous LBLOCA analyses, based on the TAC03 fuel performance code, accounted for the differences by the application of a gap gaseous conductance multiplier. The multiplier, which was held constant throughout the transient, forces the initial fuel temperature prediction of RELAP5/MOD2-B&W to match the fuel performance code prediction plus uncertainty.
An evaluation of the RELAP5/MOD2-B&W and COPERNIC gap conductance models was performed to understand the differences between the models and to determine whether the application of a constant gap gaseous conductance multiplier (determined at steady-state) remained the appropriate method for accounting for the differences between the models and for the uncertainty adjustment of the initial fuel temperatures. Figure Q25-1 illustrates the differences between the RELAP5/MOD2-B&W and COPERNIC gap gaseous conductance models. The figure shows the multiplier on the RELAP5/MQD2-B&W term that would be necessary for it to match the COPERNIC prediction as a function of steady-state gap thickness. The gap thickness effectively translates to the inverse of time-in-life, where open gap conditions exist at BOL and the gap closes and contact pressures develop with increasing burnup.
53
The results of the evaluation determined that RELAP5/MOD2-B&W and COPERNIC provide similar gap conductance results when the gap thickness is relatively large.
However, there was a noticeable difference in the gap conductance when the gap is small.
The accounting of gaseous conductance for gas space between rough surfaces in contact differs between the two codes. Although a gaseous conductance multiplier would allow RELAP5IMOD2-B&W to generate an initialization that matched the uncertainty-adjusted COPERNIC fuel temperatures, the multiplier value would be large for small gaps and applicable only so long as the gap remains small.
Figure Q25-2 demonstrates the transient gap thickness for LBLOCAs initialized at BOL and 45,000 MWd/MtU. For BOL, the gap is initially open and the increased transient gap does not significantly alter the gaseous conductance. A multiplier of between one and two could be applied without significantly affecting the transient simulation. However, for exposed fuel, the initialization multiplier based on the gaseous conductance model may be as high as six and would only be reduced to between two and three by application of the COPERNIC fuel pellet temperature uncertainties. Such a multiplier would quickly become inappropriate as the gap opens during the transient. Because RELAP5IMOD2-B&W does not have the ability to modify the gap gaseous conductance multiplier during the transient, and it is apparent that the multiplier should be less than two after about five seconds, the gaseous conductance multiplier approach was deemed inappropriate for COPERNIC-based LOCA calculations.
RELAP5/MOD2-B&W does have the capability to directly specify the initial fuel rod temperatures independent of the gap conductance. It is, therefore, possible to force the initial heat structure temperatures to the correct values, albeit by giving up a strict steady-state configuration. To determine the effects of starting the core in a non-steady-state condition, a study of several fuel pins with differing gap coefficients was performed.
LOCA simulations with multiple hot fuel pins, each with the same initial fuel temperature distribution (input specified), but with gaseous conductance multipliers varying from 0.5 to 2.0 were run. The results, Figure Q25-3, demonstrated timing differences in cladding heating and cooling rates, particularly in the first few seconds of the transient. However, the overall cladding and fuel temperature trends were preserved and no significant peak cladding temperature differences were noted. The initial heatup of the cladding and cooldown of the fuel pellet occurred quicker with a high gaseous conductance multiplier.
For reduced gaseous conductance, the opposite was true. After the initial heatup, however, the offset of the cladding and fuel temperatures is aligned to compensate for the differences in the gap conductance and the cladding temperature response are thereafter consistent in both timing and magnitude. Because the fuel energy decrease is delayed for the lower gap conductance, fuel temperatures tend to remain higher during the refill and reflood portions of the LBLOCA, resulting in a tendency for a slightly higher cladding temperature during this phase. Furthermore, because the cladding temperature response is, for the most part, consistent, it can be inferred that the core energy transmitted to the reactor system, which is initialized at steady-state conditions for the plant power, is consistent and that there is not a significant effect on the evolution of the remainder of the primary system during the LOCA transient. Therefore, because Figure Q25-2 shows that the gap opens quickly during a LBLOCA and Figure Q25-1 shows that there is little difference between the gaseous conductance of RELAP5/MOD2-B&W and COPERNIC for open gaps, the best solution is to apply no gaseous conductance multiplier (i.e. a factor 54
of 1.0).
In conclusion, the system model in the MOX demonstration cases was initialized to steady state at the desired peaking conditions and the initial fuel temperatures were set to the COPERNIC-predicted temperatures with appropriate uncertainties added. The method ensures an appropriate specification of the initial fuel stored energy and a proper calculation of the gap conductance during a LBLOCA transient.
Figure Q25-1 Multipliers on RELAP5 to Match COPERNIC Gap Thermal Model 6.0 5.0 1 4.0 LBLOCA Transient Gap 8 Size Near Peak Power 3.0 and PCT Node after -5 seconds.
S0 I -
S I 2.0 I X-
- 1.0 _ v
- 0n-. I 0 25 50 75 100 125 150 175 200 Hot Mechanical Gap (micron) 55
Figure Q25-2 LBLOCA Transient Hot Mechanical Gap Sizes 250
-48BOL --- 45 GWd/mtU LO200 -- - - - - - - -- - - - - - - - -- - - - - - - -- - - - - - - -
200
.50 co -
0 0 5 10 15 20 25 30 Time after LBLOCA Initiation (sec) 56
Figure Q25-3 LBLOCA Transient Cladding Temperatures at PCT Location TfIME (a)
(Mg = Gaseous Conductance Multiplier)
- 26. Provide the basis for assuming that the uncertainty distribution for COPERNIC is a normal distribution.
Response
The actual assumption was that the COPERNIC uncertainty distribution was approximately normal. This assumption and the basis for this are explained in the response to Reactor Systems Question 24.
- 27. Please provide the basis for the COPERNIC temperature adjustments for core initialization in section 3.7.1.1.4. Additionally, please provide the basis for why the TAC03 temperature predictions are reasonable for application to COPERNIC predictions.
Response
TAC03 temperature predictions have no application to COPERNIC predictions. What was involved in the fuel temperature initialization of the LOCA core simulation was that the relative uncertainty for a specific region of the core, originally developed based on the TACO uncertainty distribution, was applied to the COPERNIC fuel temperature prediction. The application of the same relative uncertainty and the basis for it are 57
explained in the response to Reactor Systems Question 24.
- 28. In sub-section 3.7.1.6, the subject of mixed cores is discussed. In the middle of the paragraph it is stated that the MOX LTA pressure drop is less than four percent lower than the pressure drop for a resident Westinghouse fuel assembly at design flow rates.
Please provide additional detail on the cause of this pressure drop difference, how it was calculated, and the impact including the consequences of this pressure drop. Also, please provide the design flow rate used for this analysis.
Response (Previously submitted October 3, 2003)
The pressure drop difference between the resident Westinghouse Robust Fuel Assembly (RFA) fuel and the MOX fuel lead assemblies is due to mechanical design differences in the grids and the top and bottom nozzles of the fuel assemblies. Even though the rod geometry, pitch, and axial grid locations are the same, unique design differences in the grids and nozzles themselves cause differences in hydraulic resistance. This overall difference was calculated by evaluating full core RFA and full core MOX models with the VIPRE-01 thermal-hydraulic code and comparing the overall calculated Ap. The code represents these hydraulic differences by means of vendor-provided form loss coefficients for each grid design, top, and bottom nozzles. The design flow rate for these evaluations was the current Technical Specification minimum flow rate of 390,000 gpm.
The impact of this difference in pressure drop is flow redistribution between fuel types in a mixed core environment. This redistribution varies with axial elevation in the core as a direct effect of the difference in local grid form loss coefficients. The consequences of this pressure drop difference result in the need to account for this flow redistribution in the analyses of fuel assembly lift, departure from nucleate boiling ratio (DNBR) in steady state and transient analyses, and fuel assembly performance issues such as maximum allowable crossflow. Flow redistribution is accounted for in these analyses by modeling the hydraulic differences directly in a conservative representation of the mixed core fuel assembly geometry.
- 29. The staff presumes that a mixed core analysis will be performed to account for the use of four MOX LTAs in the core. Therefore, provide the mixed core penalty that was calculated. If a mixed core calculation was not performed, provide a technical justification for not performing the analysis.
Response (Previously submitted October 3, 2003)
The mixed core MOX fuel lead assembly DNBR penalty is explicitly calculated for the entire range of conditions analyzed in a reload cycle. With the currently licensed Duke Power analysis methodology, maximum allowable radial peaking limits are calculated for a range of axial peak locations and magnitudes as described in DPC-NE-2004P-A. This family of peaking limits is repeated for the various sets of reactor statepoints (power level, pressure, temperature, and flow) analyzed to support cycle reload analyses. This entire set of limits is used to represent the limiting fuel assembly in the core.
To model the mixed core, a bounding model of a single high powered MOX fuel assembly at the center of the core surrounded by a remaining core of resident Westinghouse RFA fuel assemblies was used to calculate the explicit peaking limits. This model contained 58