ML033210345
| ML033210345 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 11/04/2003 |
| From: | Tuckman M Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MB7863, TAC MB7864, TAC MB7865, TAC MB7866 | |
| Download: ML033210345 (119) | |
Text
Duke PawPower.
A Duke Energy Company MICHAEL S. TUCKMAN Executive Vice President Nuclear Generation Duke Power P.O. Box 1006 Charlotte, NC 28201-1006 Mailing Address:
526 South Church Street Charlotte, NC 28202 November 4, 2003 704 382 2200 704 382 4360 fax U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001
Subject:
Reference:
Catawba Nuclear Station Units 1 and 2; Docket Nos. 50-413, 50-414 McGuire Nuclear Station Units 1 and 2; Docket Nos. 50-369, 50-370 Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies NRC Letter dated July 25, 2003, Request for Additional Information Re: Mixed Oxide Lead Fuel Assemblies (TAC Nos. MB7863, 7864, 7865, 7866)
By letter dated November 3, 2003 Duke Power transmitted its proprietary response to the reference Request for Additional Information. Attached is the redacted version of this response with all proprietary information removed. A copy of the affidavit that accompanied the original submittal attesting to the proprietary nature of the information is also attached.
Inquiries on this matter should be directed to G.A. Copp at (704) 373-5620.
M. S. Tuckman attachment 4
ADC)(
www. duke-energy. corn
U. S. Nuclear Regulatory Commission November 4, 2003 Page 2 cc: Nv/o attachment R. E. Martin, NRC Senior Project Manager (0-8 G9)
U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 L. A. Reyes, Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303
AFFIDAVIT COMMONWEALTH OF VIRGINIA
)
) Ss.
CITY OF LYNCHBURG
)
I My name is Gayle F. Elliott.
am Manager, Product Licensing, for Framatome ANP ('FANP"), and as such I am authorized to execute this Affidavit.
- 2.
am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary.
am familiar with the policies established by FANP to ensure the proper application of these criteria.
- 3.
am familiar with the FANP material enclosed in Duke Power's Response to Request for Additional Information (NRC Letter dated July 25, 2003) and referred to herein as "Document." Information contained in this EW5cument has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.
- 4.
This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.
Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5.
This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in the Document be withheld from public disclosure.
- 6.
The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:
(a)
The information reveals details of FANP's research and development plans and programs or their results.
(b)
Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c)
The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FANP.
(d)
The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FANP in product optimization or marketability.
(e)
The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.
- 7.
In accordance with FANP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside FANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8.
FANP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
- 9.
The foregoing statements are true and correct to the best of my knowledge, information, and belief.
SUBSCRIBED before me this ;2i4l day of
, 2003.
see'e5_'7
a,-
Ella F. Carr-Payne NOTARY PUBLIC, STATE OF VIRGINIA MY COMMISSION EXPIRES: 8/31/05 ELLA F. CARR-PAYNE Notary Public Commorweal o Virginia I
y abso E* Aug.31.200
Attachment Response to NRC Staff Request for Additional Information dated July 25, 2003 REACTOR SYSTEMS QUESTIONS
- 1. With respect to section 3.1, provide a full description of the post-irradiation examination planned to verify the mechanical properties of the lead test assemblies (LTAs) following irradiation. Describe the test methods planned and the acceptance criteria for each test as well as the frequency of each test. Additionally, please provide the same information for the hot cell examinations that will be performed following irradiation.
Response (Previously submitted October 3, 2003)
Post-irradiation examination (PIE) of the lead assemblies will include both poolside and hot cell examinations. A description of the examinations, including test methods, acceptance criteria, and frequency of testing, is found in Section 8.5 of Reference Ql-l.
This topical report is currently under NRC review.
The purpose of the hot-cell examinations is to collect data to confirm the applicability of fuel performance models for fuel burnups greater than the limit (50,000 MWd/MThm maximum rod average) that is proposed for batch application. Therefore, specific acceptance criteria have not been set for the hot-cell examinations.
Reference Ql-l. BAW-10238(NP) Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.
- 2. With respect to section 3.5.1.1, what is the calculated shoulder gap for the LTAs following a third cycle of irradiation? How does it compare to the limit?
Response
The minimum gap was calculated to be [
] inch at a burnup of 60,000 MWd/MThm (maximum rod average). The minimum allowable shoulder gap is zero.
The shoulder gap calculation used the 95%o/95% upper confidence limit on the predicted fuel rod growth. In addition, the maximum fuel rod length and minimum nozzle-to-nozzle span, as indicated by dimensional tolerances on the drawings, were used. This choice of as-built dimensions minimizes the initial shoulder gap.
- 3. Provide the specific burnup limit that is being requested for the LTAs.
Response (Previously submitted October 3, 2003)
The requested burnup limit for the lead assemblies is 60,000 MWd/MThm for the maximum rod (axially averaged). This is greater than the 50,000 MWdIMThm maximum rod average that is proposed for batch application. Additional information on these limits is contained in Section 8.4 of Reference Q3-1.
Reference Q3-1. BAW-1 0238(NP) Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
- 4. Section 3.5.1.2 of the submittal states that the reduction in total plutonium concentration ensures that the macroscopic plutonium effects on fuel performance are bounded. Please define the subject macroscopic plutonium effects.
Response (Previously submitted October 3, 2003)
The macroscopic effects referred to are changes in the following parameters:
Fuel thermal conductivity Pellet thermal expansion Pellet thermal creep Fission gas release Fuel densification and swelling Helium gas accumulation and release Pellet radial power profile Fuel melting point Each of these parameters is affected by the amount of PuO 2 in the MOX fuel pellets. By using approximately 40% less PuO2 in weapons grade MOX fuel relative to reactor grade MOX fuel, the effects of changes to these fuel performance parameters relative to low enriched uranium (LEU) fuel will be bounded by the European experience with reactor grade MOX fuel.
These effects are also listed in Section 2.1 of Reference Q4-1, the MOX Fuel Design Report, which has been submitted to the NRC for review. Further discussion of these macroscopic effects is provided in Sections 2 and 3 of that document. The pellet radial power profile will be addressed further in the response to Reactor Systems RAI Question 7.
Reference Q4-1. BAW-10238(P), Revision I, MOX Fuel Design Report, Framatome ANP, May 2003.
2
- 5. Provide the appropriate regulatory criteria used for the parameters discussed in section 3.5.1.
Response (Previously submitted October 3, 2003)
Section 3.5.1 provides a description of MOX fuel and fuel rod design features, and is consistent with Reference Q5-1, which is currently under NRC review. Similarly, Section 3.5.2 describes the MOX fuel assembly mechanical design features. These sections provide the necessary background for the design evaluation, which is presented in Section 3.5.3 and discusses how the design satisfies the design criteria. The parameters described in Sections 3.5.1 and 3.5.2 support a fuel design that satisfies the acceptance criteria in Section 4.2 of Reference Q5-2 that pertain to pressurized water reactor fuel assembly design.
References Q5-1. BAW-10238(P), Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
Q5-2. NUREG-0800, U. S. Nuclear Regulatory Commission Standard Review Plan, Revision 2, July 1981.
- 6. Provide a statistical analysis showing that the distribution of fissile material for the weapons-grade (WG) mixed oxide (MOX) fuel is the same as the distribution of fissile material for reactor-grade (RG) MOX fuel.
Response (Previously submitted October 3, 2003)
The requested statistical analysis is not practical until after the fuel is manufactured.
However, it is not necessary to perform such an analysis in order to show that the distribution of fissile material is approximately the same for both types of MOX fuels when using the same manufacturing process.
As discussed in Section 2.3 of Reference Q6-1, the MIMAS process ensures homogeneity of the pellet. The plutonium-rich agglomerates are finely dispersed in the U02 matrix, and this distribution is not a function of isotopic content, only the final desired plutonium content. With regard to plutonium-rich agglomerate size, the relevant portions of the European (Reactor Grade) and American (Weapons Grade) Fuel Pellet Specifications are identical. This is also discussed in Section 2.3. Therefore, the distribution of plutonium-rich agglomerates is the same in reactor-grade and weapons-grade fuels.
As discussed in Section 3.3 of Reference Q6-1, the fissile plutonium contents of the agglomerates in RG and WG MOX fuel are approximately 22.5% and 19.2%,
respectively. The fissile uranium content of the agglomerates is very small and is approximately 0.2% for both RG and WG MOX fuel, assuming a 0.25 weight percent tails uranium is used for feed powder.
Since the distribution of plutonium-rich agglomerates is the same in RG and WG MOX fuels, and the fissile plutonium content of the agglomerates is approximately equal, the 3
distribution of fissile material will also be approximately the same for both types of MOX fuel of equivalent reactivity. However, note that it is common practice to adjust the master mix to U02 ratio in order to achieve the desired final plutonium content of a given rod. Each MOX fuel assembly is composed of three different plutonium content rods. For example, the three plutonium concentrations could be 2.40, 3.35, and 4.94 w/o plutonium for corner, periphery, and interior pins, respectively (see response to Reactor Systems RAI Question 18, Figure 18-2). The adjustment for these different pins, for which there is plenty of experience, dramatically changes the fissile material distribution within the pellet and far outweighs the small difference between RG and WG MOX fuel.
Reference Q6-1. BAW-10238(P), Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
- 7. With respect to section 3.5.1.2, provide additional information on how the fuel pellet radial power profile for WG MOX is bounded by the profile for RG MOX.
Response
The statement in Section 3.5.1.2 should read that the radial power profile for weapons grade MOX fuel is between the values for LEU fuel and reactor grade MOX fuel. As shown in Figures Q7-1 through Q7-3, the weapons grade MOX fuel power profile is "bounded" above and below by either LEU fuel or reactor grade MOX fuel values. These figures were generated with the APOLLO2-F neutron transport code (Reference Q7-1) using 4.27 weight percent U-235 LEU fuel, 4.94 weight percent plutonium weapons grade MOX fuel, and 8.00 weight percent plutonium reactor grade MOX fuel pins depleted to 60 GWd/MThm. Additionally, previous analyses have shown that the effect of the different radial power profiles for reactor grade and weapons grade MOX fuel have a negligible impact on important fuel performance parameters such as fission gas release and centerline temperature (Reference Q7-2).
References Q7-1. Letter, T. A. Coleman (Framatome Cogema Fuels) to Document Control Desk (NRC), Submittal of Accepted Version of Topical Report BAW-10228P, SCIENCE, GR0192.doc, December, 2000.
Q7-2. Letter, James F. Mallay (Framatome ANP) to Document Control Desk (NRC),
Partial Response to RAI on Chapter 13 of BAW-10232P, (See response to Question 1), NRC:02:021, April 26, 2002.
4
Figure Q7-1 Pellet Radial Power Profiles 0.5 GWdIT 2.4-2.2-2-
I-30 1.8
-E a, 1.6
.... WG MOX m
~
~~~~~~-
RG MOX Z 1.4 1.2 0.0 0.5 1.0 R/Ro 5
- 8. In section 3.5.3, the statement is made that the Mark-BW/MOX1 fuel assembly design meets all applicable criteria to maintain safe plant operation. Specify all of the regulatory criteria being referred to in this statement.
Response (Previously submitted October 3, 2003)
A detailed analysis of the Mark-BW/MOXI fuel assembly design is presented in Reference Q8-1. The design satisfies the acceptance criteria in the Stantdard Revieiv Plan, Section 4.2 that pertain to pressurized water reactor fuel assembly design. The acceptance criteria in the Standard Review Plan, in turn, align with the relevant requirements (regulatory criteria) in 10 CFR 50.46, General Design Criteria 10, 27, and 35, and 10 CFR 100.
The criteria in the Standard Review Plan, Section 4.2.II.A, that are not addressed are as follows:
Il.A.1 (h): This criterion applies to control rods rather than fuel.
II.A.2(c): Fretting is treated in the response to criteria in paragraph II.A. 1 (c).
II.A.2(f): This criterion applies only to boiling water reactor fuel.
I1.A.2(i): Mechanical fracturing is treated in the response to criteria in paragraph II.A.3(e).
II.A.3(c): More stringent criteria are already applied in paragraph II.A.3(a).
6
Reference Q8-1. BAW-I 0238(NP) Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
- 9. Section 3.5.3 refers to fuel rod analyses which follow previously approved methods.
Please state the methods being referred to.
Response (Previously submitted October 3, 2003)
Approved methods used for analysis of MOX fuel are listed below. A more detailed discussion of the methods used for analysis of MOX fuel is provided in Reference Q9-1.
The fuel rod cladding stress and cladding fatigue were analyzed using approved methodologies for MS'TM cladding that are described in Reference Q9-2. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.
Cladding corrosion was analyzed using the COPERNIC code with models approved for predicting M5S'I cladding oxide thickness as described in Reference Q9-2. This methodology is applicable to both LEU and MOX fuel because it is independent of the pellet type.
Axial growth calculations for fuel rods and fuel assemblies also used the methodology that was approved in Reference Q9-2.
The hydraulic forces for assembly liftoff calculations were determined using the NRC approved LYNXT code described in Reference Q9-3. This methodology is applicable to both LEU and MOX fuel because it is independent of the pellet type.
Fuel rod end-of-life internal pressure and centerline fuel melt temperatures were evaluated according to the methods and criteria presented in Reference Q9-4 for MOX fuel rods.
Approval of Reference Q9-4 is pending. Revision 0 of this topical report has been approved for use with LEU fuel. A consistent methodology is used to describe the variations in properties between LEU and MOX fuel.
The fuel rod was analyzed for creep collapse using approved methods described in Reference Q9-5. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.
Overheating of cladding was analyzed with the approved BWU-N and BWU-Z critical heat flux correlations, which are described in References Q9-6 and Q9-7. This method is applicable to both LEU and MOX fuel because it is independent of the pellet type.
Fuel enthalpy during a rapid insertion of reactivity (control rod ejection) is calculated using the SIMULATE-3K MOX computer code as described in Section 3.7.2.4 of to the Duke Power MOX fuel lead assembly license amendment request dated February 27, 2003. SIMULATE-3K MOX is not yet an approved methodology for 7
MOX fuel, although the similar code SIMULATE-3K has been approved for application to rod ejection analyses in low enriched uranium fuel cores (Reference Q9-8).
LOCA calculations for cladding rupture, cladding embrittlement, and fuel rod ballooning used approved methods described Reference Q9-2. These methods are applicable to both LEU and MOX fuel because they are independent of the pellet type.
The axial and horizontal faulted analysis methodologies (for external forces) are consistent with the approved methodologies described in References Q9-9 and Q9-10, respectively.
These methods are applicable to both LEU and MOX fuel because they are independent of the pellet type. Application of these methods to the Advanced Mark-BW fuel assembly design is discussed Reference Q9-1 1.
References Q9-1. BAW-10238(NP) Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
Q9-2. BAW-1 0227P-A Revision 0, Evalhation ofAdianced Cladding and Stnrctural Material (M5) in PWVR Reactor Fuel, February 2000.
Q9-3. BAW-10156-A Revision 1, LYNXT: Core Transient Thernmal-Hydrauilic Program, August 1993.
Q9-4. BAW-1 0231 P Revision 2, COPERNIC Fuel Rod Design Computer Code, July 2000.
Q9-5. BAW-1 0084-A Revision 3, Program to Determine In-Reactor Performance of BW1'FC Fuel Cladding Creep Collapse, July 1995.
Q9-6. BAW-10199P-A Revision 0, The B JVU Critical Heat Flux Correlations, December 1994.
Q9-7. BAW-10199P-A Addendum 2, Application of the BJU-Z CHF Correlation to the Mark-B Wi 7 Fuel Design with Mid-Span Mixing Grids, June 2002.
Q9-8. DPC-NE-2009-P-A, Duke Power Company Westinghouse Fuel Transition Topical Report, September 1999.
Q9-9. BAW-1 01 33P-A Revision 0, Mark-C Fuel Assembly LOCA-Seismic Analysis, June 1986.
Q9-10. BAW-10133P-A Revision 1, Addendum 1 and Addendum 2, Mark-C Fuel Assembly LOCA-Seisnic Analyses, October 2000.
Q9-11. BAW-10239(NP) Revision 0, Advanced Mark-B WFuel Assembly Mechanical Design Topical Report, March 2002.
8
- 10. Section 3.5.3 states that the Mark-BNV/MOXI design preserves the original plant licensing bases for all reactor internal components. Specify the components and the bases for them that are being referred to in this statement. Also explain how the MARK-BNV/MOX1 design preserves them.
Response (Previously submitted October 3, 2003)
The reactor internal components that are affected by the fuel assembly design are the upper and lower core plates. The licensing basis is preserved through ensuring that the criteria established in Reference Q10-1 are met. These criteria were developed from the acceptance criteria of the Standard Review Plan. The axial growth and assembly liftoff criteria in Reference Q10-1 must be met during the assembly lifetime. In addition, the fuel assembly must meet the following criteria for externally applied forces:
a) Operating basis earthquake: Allow continued safe operation of the fuel assembly following an event by ensuring the fuel assembly components do not violate their dimensional requirements.
b) Safe shutdown earthquake: Ensure safe shutdown of the reactor by maintaining the overall structural integrity of the fuel assemblies, control rod insertability, and a coolable geometry within the deformation limits consistent with the emergency core cooling system and safety analysis.
c) LOCA or safe shutdown earthquake plus LOCA: Ensure safe shutdown of the reactor by maintaining the overall structural integrity of the fuel assemblies and a coolable geometry within deformation limits consistent with the emergency core cooling system and safety analysis.
The Mark-BW/MOXI design preserves the design basis because the growth models, nominal dimensions, and tolerances for the holddown springs, nozzles, and guide thimble assemblies are the same as those given in Reference QlO-2.
References QIO-1. BAW-10238(NP) Revision 1, MOXFtuel Design Report, Framatome ANP, May 2003.
QI0-2. BAW-10239(NP) Revision 0, Advanced Mark-B WFtrel Assemtbly Mechanical Design Topical Report, March 2002.
- 11. Please respond to the following items related to section 3.6.1, so that the Nuclear Regulatory Commission (NRC) staff may evaluate the conclusion that the vessel fluence increase is limited:
(A) Identify the location of the LTAs.
(B) Identify the existing peak fluence azimuthal location of the vessel.
Response (A) (Previously submitted October 3, 2003)
The MOX fuel lead assembly license amendment request is not tied to a specific unit and cycle. Current plans are to insert the lead assemblies in Catawba Unit I Cycle 16 (CIC16), which will start up in late spring 2005. The C1C16 core design will not be finalized until the first quarter 2004. The location of the lead assemblies will be 9
influenced by numerous design considerations, and it is not feasible to commit to the location at this time. Duke has made a commitment (Reference Qi 1-l)to the Nuclear Regulatory Commission to place at least two of the four lead assemblies in core locations that contain incore flux mapping instrumentation. Core location C-08 (and its symmetrical locations) is a potentially desirable location because it is the only core location that will (i) allow full instrumentation for all four assemblies and (ii) maintain an eighth core symmetric arrangement. Peripheral core locations would not be desirable because the lead assemblies would not be subjected to sufficient fuel duty to achieve burnups representative of MOX fuel batch implementation. Duke has also made a commitment (Reference QI 1-1) not to place the lead assembly in a core location with a control rod during the first cycle of use.
Figure QI 1-1 depicts the Catawba one-quarter core geometry. The locations shaded in gray and red show typical fresh fuel placement in recent Catawba fuel cycles. The desirable lead assembly location C-08 (and its symmetrical location H-13) is shown in red.
10
Figure Q1 1-1 Representative MOX Fuel Lead Assembly Core Design H
G F
E D
C B
A 8
9 10 11 12 13 14 15 Fee-edee
,LEU'
~.
4 LEV'j_
LEU~~i LEU r-'
EU!e FEe r~~d tF
.0efr
,,,, ' q- 4 _
i,
l- -%-LEU -
- i z 4z; '.A t
, I
-'
i I.
.I T IWd T__'
I FedIIw*
+ @1 WS It F
E
9.
.9-,-
.9 Response (B) (Previously submitted October 3, 2003)
For Catawba Units 1 and 2, the peak fast neutron fluence at the pressure vessel cladding/base metal interface is at the 300 azimuth relative to the core cardinal axis.
Catawvba cores are designed with eighth-core symmetry. Accordingly, relative to the diagram shown in Figure Q1 1-1, the core cardinal axis would be through row eight.
Figures Q1 1-2 and Q1 1-3 show the relative fast neutron flux differences between a core with four MOX fuel lead assemblies and an all LEU core at beginning of cycle and end of cycle, respectively (This information was not specifically requested but it shows the negligible impact of four MOX fuel lead assemblies on core exterior fast flux, which control vessel fluence).
Reference Q1 1-1. Tuckman, M.S., June 26,2003, Letter to U.S. Nuclear Regulatory Commission, Physics Testing Program in Support of Topical Report DPC-NE-1005P Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, August 2001.
11
Figure Q1 1-2 Fast Neutron Flux Comparison All LEU Core vs Core with Four MOX Fuel Assemblies (Beginning of Cycle)
Flux above 0.625 Mev HFP, 4 EFPD, ARO, Nominal Conditions H
G F
E D
C B
A 8
9 3.619 3.561 13.6521 3.654 3.688 3.7091 2.970 1.429 3.607 3.551 3.649 3.670 3.761 3.948 3.050 1.452
-0.3
-0.3
-0.1 0.4 2.0 6.4 2.7 1.6 3.597 3.587
-0.3 3.565 3.560
-0.1 3.626 3.635 0.2 3.604 3.644 1.1 3.442 3.515 2.1 3.121 3.170 1.6 1.629 1.646 1.0 10 3.587 3.581
-0.2 3.555 3.552
-0.1 3.588 3.593 0.1 3.582 3.594 0.3 3.260 3.271 0.3 1.652 1.657 0.3 A _9.
_9 A
A9 A
.9__-
11 3.584 3.571
-0.4 3.680 3.664
-0.4 3.664 3.649
-0.4 3.055 3.044
-0.4 1.220 1.217
-0.2 A_ A9 A_ A _
A.A-12 3.975 3.947
-0.7 3.678 3.650
-0.8 3.006 2.979
-0.9 2.240 2.223
-0.8 1.325 1.312
% relative diff Figure Q1 1-3 Fast Neutron Flux Comparison All LEU Core vs Core with Four MOX Fuel Assemblies (End of Cycle)
Flux above 0.625 Mev HFP, 495 EFPD, ARO, Nominal Conditions H
G F
E D
C B
A 8
9 3.857 3.983 r 3.679 3.434 3.534 3.667 3.072 1.651 3.872 3.997 3.692 3.438 3.514 3.647 3.047 1.648 0.4 0.4 0.4 0.1
-0.6
-0.5
-0.8
-0.2
_~ ~ ~~~
.4 3.0 3.4 3.41 1.876 3.811 3.825 0.4 3.916 3.929 0.3 3.644 3.652 0.2 3.804 3.798
-0.2 3.540 3.520
-0.6 3.410 3.400
-0.3 1.876 1.877 0.1
_~ ~
~.
-0.
-0.
-0____
__.3 0.1_
10 3.827 3.841 0.4 3.968 3.981 0.3 3.782 3.790 0.2 3.909 3.916 0.2 3.548 3.555 0.2 1.872 1.878 0.3 I
I 11 3.830 3.844 0.4 3.958 3.973 0.4 3.605 3.619 0.4 3.052 3.065 0.4 1.345 1.352 0.5 l
bI
- l l
l 12 3.790 3.807 0.4 3.524 3.541 0.5 2.872 2.887 0.5 2.196 2.208 0.5 1.356 1.364 0.6 13 No MOX 4 MOX
% relative diff Note: Figures indicate fast neutron flux scaled by 1x1 014, in units of neutrons/cm 2/sec 12
- 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 I OCFR 50.46 (b) were established for light water reactors fueled with U02 pellets within cylindrical Zircaloy cladding. The MOX fuel lead assemblies have M5S'~l 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 Standard Review Plait (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 cal/gm, 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 Q12-4, 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-1).
- 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. M5TM 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 Q12-7) 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-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Duke Power Company, December 2000.
Q12-2. NUREG-0800, U. S. Nuclear Regulatory Commission Standard Review 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 Data for Reactivity-Insertion Accidents, Nuclear Safety, Volume 37, No. 4, October-December 1996.
Q1 2-4. EPRI Technical Report 1002865, Topical Report on Reactivity Initiated Accident: Bases for RIA Fuel and Core Coolability Criteria, June 2002 (currently under NRC review).
Q12-5. BAW-10238(P), Revision 1, MOXFuiel Design Report, Framatome ANP, May 2003 (currently under NRC review).
Q12-6. DPC-NE-1 005P, Duke Power Nuclear Design Methodology Using CASMO-4/SIMULA TE-3 MOX, August 2001 (currently under NRC review).
Q12-7. DPC-NE-2009-P-A, Revision 2, Duke Power Company Westingholuse Fuel Transition Report, December 2002.
Q 12-8. 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 I OCFR 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 I OCFR 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, 'he concert 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 O.:10CFR. * "
- v..50.46 (b) E
' >Criteria'.
- - ApplicabilitytoMOX Fuel Lead Assemblies Peak Clad Temperature
< 2200 OF 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 M5T cladding.
The 2200 OF criterion has been approved by the NRC as applicable to M5T' 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 that MST behavior during LOCA conditions is equivalent to or superior to Zircaloy and is documented in BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5T) 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 UO2 or MOX fuel pellets.
This criterion is fully applicable to the MOX fuel lead assemblies.
17% Local Oxidation 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 MST cladding.
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 that M5TM behavior during LOCA conditions is equivalent to or superior to Zircaloy 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 1 % Core-unrelated to the core fuel and cladding so long as the hydrogen produced per wide percent cladding reacted is unchanged. Because the reaction for both M5 M and Oxidation Zircaloy is between zirconium and oxygen, the hydrogen produced per reaction percent is the same for both materials. The criterion is unaffected by the use of M5 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
This criterion controls the availability of long-term cooling systems and core Long-term 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 Coolingassemblies.
18
Table Q1 2-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies TransientlAccident
-Acceptance Criteria' Description 6.2.1.3 LOCA Mass and Energy Containment design margin is maintained.
Release and Containment Environmental qualification of the safety related equipment inside Pressure/Temperature Response containment is not compromised.
6.2.1.4 Secondary System Pipe Containment design margin is maintained.
Ruptures and Containment Environmental qualification of the safety related equipment inside Pressure/Temperature Response containment is not compromised.
15.1.1 Feedwater System Malfunctions that Result in a Bounded by excessive increase in secondary steam flow analysis Reduction in Feedwater in Section 15.1.2 and same criteria apply.
Temperature Peak RCS pressure remains below 110% of the design limit 15.1.2 Feedwater System
(<2750 psia)
Malfunction 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) 15.1.3r EcSteam Flow a
Fuel cladding integrity shall be maintained by ensuring that the Secondary Steam 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 15.1.4 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 95/95 DNBR
. Steam SystemPipinlimit based on an acceptable DNBR correlation. If the DNBR falls 151.5 Steam System Pipingbelow these values, fuel failure must be assumed for all rods that do 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 1 OCFR100.
19
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Transient/Accident Acpac~iei
- escri ti n
~
cceptance Criteria:
-Description 15.2.1 Steam Pressure Regulator Not applicable, there are no pressure regulators in the McGuire or alfunction or Failure That Results Catawba plants whose failure or malfunction could cause a steam In 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) 15.2.3 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 0o Bounded by turbine trip analysis in Section 15.2.3 and same Vacuum 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-Emnergency
(<2750 psia)
C Power to theNStation-Auxiliesy 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)
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) for low probability events.
15.2.8 Feedwater System Pipe Fuel cladding integrity shall be maintained by ensuring that the reak 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 psia)
Reactor 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 012-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies TransientiAcciden-Descriptance Criteria,,
Peak RCS pressure remains below 110% of the design limit 15.3.2 Complete Loss of Forced
(<2750 psia)
Reactor 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 10CFR100 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) ontrol Assembly Bank Fuel cladding integrity shall be maintained by ensuring that the Withdrawal From a Subcritical or calculated DNB ratio remains above the 95/95 DNBR limit based Low Power Startup Condition on an acceptable DNBR correlation.
o 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) 15.4.2 Uncontrolled Bank Cluster 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% of the design limit 15.4.3 Rod Cluster Control
(< 2750 psia) 15.4.3 RodClusteraCon(Systrol Fuel cladding integrity shall be maintained by ensuring that the Assembly Misoperation (System calculated DNB ratio remains above the 95195 DNBR limit based Malfuntonp rOeatrErr 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 95/95 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 TransienVAccident Acceptance Criteria Description 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 Incorrect Temperature calculated DNB ratio remains above the 95/95 DNBR limit based on an acceptable DNBR correlation.
15.4.6 Chemical and Volume 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 oolant on an acceptable DNBR correlation.
15.4.7 Inadvertent Loading and Any activity release must be such that the calculated doses at the Operation of a Fuel Assembly in site boundary are a small fraction of the 10CFR100 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 10CFR1 00.
See provisional cal/gm acceptance criteria attached.
Peak RCS pressure remains below 110% of the design limit 15.5.1 Inadvertent Operation of
(<2750 psia)
Emergency Core Cooling System 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.
15.5.2 Chemical and Volume Bounded by inadvertent operation of emergency core cooling Increases Reactor Coolant system during power operation analysis in Section 15.5.1 and Inventory same 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 limit based on an acceptable DNBR correlation.
15.6.2 Break In Instrument Line or ther 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 1 OCFR1 00 guidelines.
ontainment 22
Table Q12-2 Acceptance Criteria for Non-LOCA Transients/Accidents with MOX Fuel Lead Assemblies Transient/Accident.--i; ;--.-...........-.X.:;
.ra
-s.entl-cc-de.t-'Acceptance Criteria Description-:.
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 10CFRI00 guidelines.
23
Table Q12-3 Regulatory Dose Criteria For Accidents with MOX Fuel Lead Assemblies t t.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
ii.
Accident Source Tern,:
Alt~~~~-Kernative Ac~derit" Clai'c A
Sou; T
Reference Sr Reference
~ ~~~~~~Offite Do e (E B ~ dLPZ) 300 Rem Thyroid RG' 1.195 RG 1.183 LOCA Rem roi 10CFR100.11 25 Rem TEDE R50.18 w c r
o s2.5 Rem WB SRP1 15.6.5 App. ACFR5.67 Steam Generator Tube Rupture 300 Rem Thyroid 10CFR100.11/
25 Rem TEDE RG 1.183 with fuel failure or pre-incident Iodine spike 25 Rem WB SRP 15.1.3 25 A
1
.18 SRP 15.195 Steam GeneratorTube Rupture 30 Rem Thyroid RG 1.195 RG 1.183 with concurrent Iodine spike 2.5 Rem WB I0CFR100.1
.5 2.5 Rem TEDE AOCFR50.67 SRP 15.6.3 Main Steam Line Break 5 Rem Tr RG 1.195 RG 1.183 with fuel failure or preincident Iodine spike 35 Rem Thyroid 1CFR1
.11 25 Rem TEDE IOCFR51.67 25 Rem WB
~SRP 15.1.5 App.A Main Steam Line Break 30 Rem Thyroid RG 1.195
. 2.5 Rem TEDE RG 1.183 with concurrent Iodine spike 2.5 Rem WB IOCRP15.1. App A 2_5__Rem______________1_183 Locked Rotor Accident 30 Rem Thyroid RG 1.195 2
Rem 1:>8 2.5 Rem WB SRP 15.3.3 2.5 Rem TEDE 1
.18 Rod Ejection Accident 75 Rem Thyroid RG 1.19563ReTEEG118 6.3 Rem WB2 SRP 15.4.8 App A 6.ReTEEG118 75 Rem Thyroid RG 1.195 Fuel Handling Accident 6.3 Rem WB2 SRP 15.7.4 6.3 Rem TEDE RG 1.183 RG~~~~~~~~~~~
1.9RpcfialGtte1httisciein a
euedi1iuofteoeinte5 50 Rem Thy'roid 3OCRG O
RG 1.18 All 5 Rem WB3 Appendix/
5 Rem TEDE RG1 O 5.183 50 Rem skin Appndi 9IOF506 WB= Whole body, RG=Regulatory Guide, SRP= Standard Review Plan Where a conflict exists between SRP and RG 1. 195 on the whole body dose lim~it for a particular accident, the more current guidance is shown.
RG 1. 195 specifically states that this criterion may be used in lieu of the one in the SRP.
2 3
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 fi.
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 Q13-1 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 - 1 OCFR50.46 Appendix K requirement.)
Cards 10000S80-1000OS99: 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 ICCCG801-lCCCG899: Left Boundary Heat Structure Cards Cards I CCCG901 -1 CCCG999: 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 M5Th1 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 NMSSV TSV
%ISIV n
l K A._T1 A
=
I T
r 1
780 t
6;0 770 L
h J
760 765 750 75 740 745 M
rsV t r 1 I
~~~~~~~~~~~~~~410-I 125-9 125-8 410-2 733-3 7243-125-10 125-7 410-3 733-2 733.2 72-125-12 125-5 410-5 125 13 125-4 410-6 733-1 733-1 725-2 125-14 125-3 410-7 125IS
_ 125-2 410-8 725-3 135_
{
170 c LoopAccumubtr W +
SarcW~~~~~~~~~~~~~~~aft l
150 Re"a~lHeat Tnpic Loop~~~~~emva
.1L2J L!Lr-w.
,~~~~~~~
665~~~~~6 655
~~~~~650 N1F~~IV 645 640 624 633-3 225-225-9 633-3 X 225-7 225-10 225-6 225-11 625-I 633-2 633-2 225-5 225-12 22.54 225-13 625-2 633-1 633-1 225-3 225-14 22-2 225-I5 625-3 630-3 _
225-1 225-16 630 Reactor Vessel..
r
_ * = +_
a N ~~~~~~~~~~235 240 Containmente 255 245 250O Single Loop 27
Figure Q13-2 Core Noding Diagram Upper Head Upper Plenum Outlet Annulus CV 170 CV 280
. I.x*CV 100 CV 200 Lowver Plenum 28
- 14. Provide the refcerence 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 followving 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 1 00F 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 Q14-3. These modifications are described in Section 3.7.1.2 of 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 370F). 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-380F).
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 KBU 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 KEBJ 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 I OCFR50.46 are met.
References Q14-1. WCAP-12945P-A, Volume I Revision 2 and Volumes 2-5 Revision 1, Code QualiJi cation Document for Best-Estimate Loss of Coolant Analysis, March 1998.
Q14-2. WCAP-100564P-A, Westinghouse Small Break ECCS Evaluation Model using the NOTR UMP Code, August 1985.
Q14-3. BAW-101 68P-A, Revision 3, RSG LOCA - B WNTLoss-of-Coolant Accident Evaluation Modelfor Recirculating Steanz Generator Plants, 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 J0%
Tube Plugging
--TIL ~
Elevation',
z'.
F
-(GWd/MThrh),
3(a u i l
F PT (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 1.458 2.332 2019.5 31
Figure Q14-1 MOX Fuel Lead Assembly Total Core Peaking Factor 2.5 2.4 2.3 IL 2.2 cm r-c,M.
c 2.1 C) 0
-O 2 1.9 1.8 1.7
~
~
~~~~~~~.; 8.5656 ft
-^-10.2756 ft 2-44-- 12 f-
~
~
~
~
~
~~
~~~~~
0 10 20 30 40 50 60 Burnup (GWdIMThm) 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 Qualification Document for Best-Estinmate Loss of Coolant Analysis, 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 Q17-1). 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/MThm.
Reference Q17-1. BAW-10231P Revision 2, COPERNICFutel Rod Design Conmputer 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
reactivity were placed in a checker board arrangement to simulate face adjacent MOX and LEU fiel assemb'i-s (see Figure Q18-1). Two cases were considered - one using MOX fuel with weapons grade (WG) plutonium isotopics, and another using MOX fuel with reactor grade (RG) plutonium isotopics. The lead assemblies will be WG MOX fuel; the RG MOX fuel cases are included for illustration purposes. Each case is examined at two burnup conditions: one with all fresh fuel, and the other with 20 GWd/MThm bumup on each assembly (typical of the end of one cycle of operation).
The MOX fuel assemblies used plutonium concentration zoning as shown in Figure Q1 8-2. This is standard practice in European pressurized water reactors that use MOX fuel. Low plutonium concentrations in the corner and peripheral pins serve to flatten the power distribution in the MOX fuel assemblies. The LEU fuel assemblies were a uniform lattice of 4.27 weight percent (w/o) 235U. The MOX and LEU fuel share the same dimensional characteristics (e.g., pellet diameter, cladding inner and outer diameter, and lattice pitch).
Figures Qi 8-3 through Q1 8-6 show how neutron flux and power change in a single row of pin cells traversing the MOX/LEU fuel assembly interface. Pin cell locations -9 (minus nine) thru -l (minus one) are LEU fuel pins. Pin cell locations 1 thru 9 are MOX fuel pins. The zero location on the x-axis of each graph corresponds to the center of the inter-assembly gap. Fast flux (>.625 MeV), thermal flux (< 0.625 MeV), and power are each normalized to the colorset average value.
Figures Q1 8-3 and Q1 8-4 show the results for a WG MOX/LEU colorset at burnups of 0 GWd/MThm and 20,000 GWd/MThm, respectively. Figures QI8-5 and QI8-6 show the corresponding results for a RG MOX/LEU colorset. The results are qualitatively similar between MOX fuel types and between burnups, with key points discussed below.
a) The fast flux is nearly uniform across the assemblies. Consistent with the results of the core simulations in the response to Reactor Systems RAI Question I1, the fast flux in MOX fuel is only slightly higher than the fast flux in LEU fuel.
b) There is a steep thermal flux gradient between the assemblies. The flux gradient results from the fact that the thermal neutron absorption cross-section in plutonium is larger than in uranium.
c) The thermal flux gradient is more pronounced for the RG MOX fuel case than for the WG MOX fuel case. The higher overall plutonium concentration in the RG MOX fuel assembly (7.07 w/o) depresses the thermal flux more than the WG MOX fuel assembly with only 4.37 w/o plutonium. Therefore, the thermal flux gradient from the LEU to the MOX fuel assembly is steeper in the RG MOX fuel case. It is not unexpected to find that WG MOX fuel behavior falls between that of LEU fuel and RG MOX fuel; this is consistent with studies of other neutronic characteristics of MOX fuel (Reference Q18-1).
d) Most importantly, the plutonium concentration zoning in the MOX fuel assemblies is effective in producing a relatively flat power profile across the LEU and MOX fuel assemblies. This conclusion is valid for both WG MOX fuel and RG MOX fuel. The edge and corner pins MOX pins see a much higher thermal flux than the interior fuel pins, but the lower plutonium concentration in the edge and corner pins makes the fission rate and power about the same as the interior pins. The response to a LOCA is driven by the pin power profile, not by the pin flux profile. Therefore, the neutron 34
flux behavior at the interface of these MOX and LEU fuel assemblies should have no significant impac-' on the cladding temperature response following a loss of coolant accident.
Reference Q18-1. BAW-10238(P), Revision 1, MOXFzuel Design Report, Section 3.1, Framatome ANP, May 2003.
35
Figure Q18-1 MOX Fuel and LEU Fuel Colorsets Reactor Grade MOX Weapons Grade MOX LEU 4.27 wlo U-235
",.'WG MOXW,,
I ~z-,WGM A 7.w o Pu.
'4
'4 =
K......;W.
LEU 4.27 w/o U-235 Figure Q18-2 MOX Fuel Assembly Zoning 32 2 2 2 2 2 2 2
1 2
2 2
2
'3 3 2 2
1 1
5 1 2
'. r, 1 '
2 2
21 3
2 2
F 1
I 1
I.
I r
I I
.112 2
2 1.~ ; 1 xI
't1.,1 j,1 1'
1
~1 I ;I V
,1 l.__A 2
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.1
,11..-
1.1 I
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.1 i
' 1 1.1 1
ii I 1-.
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I II 2
2 2i j 1 1 1.r 1.
'i.r
.1.
_1 1 2 2 2
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i
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.1 I
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l1 I) ii M14 2
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. -1
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__'1 1
2 2
2 1 I I
I 1 I
I 1
1 1,
2 2
3 2
2 1
.1 2
i 2
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1 2
2 3i 3; 3" 2
2 2
2 2
2 2
2 2
2 2
2 2
3'.
Pin I
2 3
Number 176 76 12 RG w/o Pu 8.00 5.42 3.89 7.07 WG w/o Pu 4.94 3.35 2.40 4.37 Total I Avg 264 36
Figure Q18-3 Weapons Grade MOX/LEU Fuel Colorset (0 GWd/MThm) 2.00 1.75 I _______'I-] _______ IiJzf;ys; I
_______ I
.,.', owerTmal Flux.I 4 3
'..- l -
t.i t.
r tS4
.-,1,t r1 O
0)
'U a)
O L_n 0
0 0
(0)
I-0) 0 1.50 1.25 1.00 0.75 0.50
+
1w V U ml K -
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..
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I 310
- "S ka
I -'
.4-4
-'.,30 I
- '"-'
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.4 t
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.6 z:-
FTUTTT71 IflGMOX'K 0.25 0.00
-10
-8
-6
-4
-2 0
2 4
6 8
10 Pin Cells Figure 018-4 Weapons Grade MOX/LEU Fuel Colorset (20 GWd/MThm) 2.00 Flux--.
Fast"lu O~~~* ;t tK N
I
'i;t- ~iM41 Ss~-
iL*Pw 1.75 4' I 7
l:
^Thermal Flux,:'
° 1.50
'1 1
C~~~~~~~~~~~~~~~~~~
0.5
° 0 7 5 s+
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s r-
- l 0 25 0275 2..,tn,
.. L E U~ t W <
.L s 'j
-aW G ;M O X 0.00 I
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-10
-8
-6
-4
-2 0
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10 Pin Cells 37
Figure Q18-5 Reactor Grade MOX/LEU Fuel Colorset (0 GWd/MThm) 2.00 1.75 al 2
1.50 I,0 I
1.25
° 1.00 0
° 0.75 Z
0.50 a:
0.25 0.00
---I I
I I
I 4
r Therhmal F
-...P,v 3,- _1*.; - I...
J...
..- I
_ _ N
_U R_
_MOX
{
-10
-8
-6
-4
-2 0
2 4
6 8
10 Pin Cells Figure Q18-6 Reactor Grade MOXILEU Fuel Colorset (20 GWd/MThm) 2.00 1.75 0
X 1.50 I,0 1.25 e)
° 1.00 0L-)
° 0.75 a) 0.50 0.25 0.00
94
'1..
p
- i*
Fast
.Tera u
t
- - N
,.,; t V~i
.',,s -,iV S, _ _ X. *.
a,
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_____L E
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li
-10
-8
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10 Pin Cells 38
- 19. Section 3.7.1.1.1 discusses a variety of neutronic parameters. Provide additional detail about the differences between LEU and MOX parameters. Please use graphs, data, or any other visual representations to help clarify the impact of these parameter differences.
Response (Previously submitted October 3, 2003)
Information on differences between LEU and MOX fuel neutronic parameters was provided in Tables 3-7 through 3-10 and Figure 3-2 in Attachment 3 to Reference Q19-1.
Additional information is provided in the responses to Reactor Systems RAI Questions 11, 18, 20, and 23.
Reference Q19-1. 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.
- 20. Also, in the second paragraph of sub-section 3.7.1.1.1, the change in the delayed neutron fraction is discussed at beginning of life. However, the behavior of the delayed neutron fraction at middle of life and end of life is not addressed. Provide a discussion on the delayed neutron factor change with respect to burnup.
Response (Previously submitted October 3, 2003)
Table Q20-1 and Figure Q20-i show the variation in core average effective delayed neutron fraction (Peff) with cycle bumup and MOX fuel loading. The introduction of four MOX fuel lead assemblies results in a minor decrease in beginning of cycle (BOC) Peff.
The impact on end of cycle (EOC) Perr is even smaller. The 40% MOX fuel core sees a more significant reduction in Deff at BOC, but the change in Defirwith bumup is much smaller than the corresponding change in a core containing all low enriched uranium (LEU). The result is that the 40% MOX fuel core has more uniform kinetics over the cycle.
39
Table Q20-1 Impact of MOX Fuel Loading and Burnup on Effective Delayed Neutron Fraction (Peff)
MOX Lead Batch MOX EFPD LEU Core' Assembly Core Core 4
6.216E-03 6.089E-03 4.827E-03 12 6.180E-03 6.060E-03 4.829E-03 25 6.125E-03 6.015E-03 4.825E-03 50 6.026E-03 5.929E-03 4.81 OE-03 100 5.852E-03 5.772E-03 4.768E-03 150 5.704E-03 5.634E-03 4.726E-03 200 5.579E-03 5.516E-03 4.686E-03 250 5.471 E-03 5.413E-03 4.650E-03 300 5.377E-03 5.322E-03 4.615E-03 350 5.292E-03 5.240E-03 4.583E-03 400 5.216E-03 5.166E-03 4.551 E-03 450 5.146E-03 5.098E-03 4.521 E-03 490 5.086E-03 5.041 E-03 4.502E-03 1 No MOX fuel assemblies 2 4 MOX fuel assemblies 3 76 MOX fuel assemblies (-40% MOX fuel core)
Figure Q20-1 Impact of MOX Fuel Loading and Burnup on Effective Delayed Neutron Fraction (Peff) 0.0065 0.0060 LU 0
U al CD 9
f!0 0.0055 0.0050 0.0045 0.0040 0
100 200 300 400 500 Cycle Exposure, EFPD 40
- 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-10231P Revision 2, COPERNIC Fuel 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 LEU Fuel (Fuel porosity of 0.0479) 9.01E 8.OE c~ 7.01E I.-
2.01E LEU ~Oa GWd/mtLJ
--- LEU @ 40 GWd/mtIU
-- 4 wti% MOX Oa0 GWd'mnthm
- <- 4 wi% NOX Ca40 GWcLmthm 0
500 1000 1500 2000 2500 3000 Fuel Temperature (F) 3500 4000 4500 5000 5500 42
Figure Q21-2 MOX and LEU Fuel Pin Temperature Profile Comparison at Loss of Coolant Accident Initiation 3500 3000 2500 -
C-2 4) 2000 -
VFuel C
FladdinFluid
~~~~~~~~~~~~~~~~~~~A _ --- -- -- -- I....,.......... I-
...... --- ------ -- j----
'j
,\\,
~
~~I i
I e----------------
3 LEU
'--------------'-------------l------
4
° -4at% NIOX z
z j
I
~
~~~~~~~I 1500 1000-500 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 -
F."I I
Cap 5
CladdIng B
Flald 3000
...........I 2000 1 000 - - - - - - - - - - - - - - -
1500 II 0
1 2
5 6
7 8
Radial Mesh Point (#)
44
Figure Q21-4 MOX and LEU Fuel Centerline Temperature Comparison for Loss of Coolant Accident 3500 3000.
e 2500-E 1 2000 w
LL.
C F 1500 1000 500
-~~~~~~LEU (Fq-2.4) 4 4wV/.MOX (Fq-2.4) 0 50 100 150 200 Time (seconds) 250 300 350 400 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) ANSIIANS-5.1-1994 standard for absorption burnup correction factors, G,,, (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 ANSIIANS-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 Product Buildup and Decay, and Associated Radiation 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 107 Seconds Reference of Q22)
Shutdown) after Shutdown)
(1)
One sigma uncertainty for 235U 2.8%
1.7 - 2.8%
ANSI/ ANS-5.1-1994, Page 14 fission product decay heat (1)
One sigma uncertainty for 238U 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 235U (MeV per Fission)
+0.5 NA ANSI/ANS-5.1-1994, Page 38 (2)
Q-sigma for 23U (MeV per Fission)
+/-1.0 NA ANSI/ANS-5.1-1994, Page 38 (2)
Q-sigma for 239Pu (MeV per
+0.7 NA ANSI/ANS-5.1-1994, Page 38 (2) 0-isigma for 241pU (MeV per
+0.7 NA ANSI/ANS-5.1-1994, Page 38 (3)
G,,a. (t) (Note 1) 2%
2.0 -18.1%
ANSI/ANS-5.1-1994, Page 26 (4) 239U decay heat one sigma 10%
NA Note 2 uncertainty (4) 239Np decay heat one sigma 15%
NA Note 2 uncertainty (4)
Decay heat for all other actinides 20%
NA Note 2 one sigma uncertainty I
I NA - Not applicable because there is no apparent time dependence of this parameter.
Note 1: Ga,(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
- 23. All operating plants must have sufficient shutdown margin at the beginning and throughout a fuel cycle. Provide the predicted shutdown margin in graphical or tabulated form for the cores including the four MOX lead test assemblies. Show that the predicted shutdown margin will meet the Technical Specification limit for each of the plants that may load the LTAs.
Response (Previously submitted October 3, 2003)
Results of quantitative Mode 1 shutdown margin calculations are presented in Table Q23-1 for a representative core consisting of 189 LEU fuel assemblies and four unirradiated MOX fuel lead assemblies. The results are compared against the same core with 193 LEU fuel assemblies. Shutdown margins are shown at beginning, middle, and end of cycle.
The calculated shutdown margins include standard conservatisms. Operation at the control rod insertion limit is assumed. Power defect is increased and control rod worth is decreased to account for uncertainty in those values. It is assumed that the highest worth control rod remains stuck out of the core, and the worth of that rod is conservatively increased.
The resulting shutdown margins are substantially greater than the current Catawba Mode 1 minimum shutdown margin operating limit of 1300 pcm. Furthermore, there are no significant adverse impacts on calculated shutdown margin due to the presence of four MOX fuel lead assemblies. The differences observed are within normal fuel cycle design variations.
As part of the standard Catawba reload design process, Duke calculates control rod worths and verifies adequate shutdown margin. This same standard reload design process will be applied to the design of the actual MOX fuel lead assembly cores, thereby ensuring that shutdown margin limits are met.
Table Q23-1 Shutdown Margins Impact of Four Fresh MOX Fuel Lead Assemblies at Catawba
- --Typp o. Core d
a Sh td M;ar -ln
)'c of Coe
-z<
Calculated ow aginpcn
.-- ;;, -1.
4EFPD
.250EFPD
.i 490 EFPD*
All LEU Fuel 3237 2498 2020 4 Fresh MOX Fuel Lead Assemblies 3255 2465 1988 49
- 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 TACO3-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 TACO3, the appropriate 95/95 confidence level for the aggregate initial temperature or stored energy of a group of eight pins requires that the TACO3 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 TACO3 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 TACO3 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 TACO3 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 TACO3 times the hot pin adjustment for COPERNIC, 3.0/11.5 times [
].
51
Reference Q24-1. Tuckman, Mv:. 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 Framatome Technologies Topical Report BA TW-10164P Revision 4, RELAP5/MOD2-B& JW, An Advanced Coniputer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis, 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 10CFR50 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/MOD2-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 RELAP5/MOD2-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 RELAP5/MOD2-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-A-
ILBLOCA Transient Gap Size Near Peak Power 9 3.0
\\
l and PCT Node after-5 seconds.
2.0-0 25 50 75 100 125 150 175 200 Hot Mechanical Gap (mkron) 55
Figure Q25-2 LBLOCA Transient Hot Mechanical Gap Sizes 250 00° 200 IE In 0
xC 150 V
I-C,
- E 100 U
C E 50 0
0 0
5 10 15 20 25 30 Time after LBLOCA Initiation (sec) 56
Figure Q25-3 LBLOCA Transient Cladding Temperatures at PCT Location I
TIME W (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 Lp. 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
the correct geometry and local form loss coefficients to represent both fuel types (see response to Reactor Systems RAI Question 28). Therefore, the mixed core peaking penalty is calculated for each unique set of conditions and the appropriate conservative limits will be applied to the lead assembly core positions in the specific cycle reload analyses. This penalty magnitude varies as a function of the axial power distribution, with the overall average penalty equal to 3% in radial peaking or 10% in DNBR, relative to a full core of Westinghouse RFA fuel assemblies.
- 30. Page 3-29 of section 3.7.2.1 lists the transients and accidents that were analyzed.
(A) Results were not provided for review with the application; therefore, submit the results of the analyses along with a discussion of each analysis and any data used for determining the impact of using MOX fuel.
(B)
Was the small break LOCA boron dilution event analyzed? If not, please provide technical justification for why it was not analyzed.
Response (A)
Specific analyses were not performed for all transients and accidents in Section 3.7.2.1.
Only the rod ejection accident was analyzed. However, each transient and accident listed in Section 3.7.2.1 was evaluated to determine if a specific analysis was needed as a result of incorporating four MOX fuel lead assemblies into the core design. The results of each of these evaluations are documented in Table Q30-1. For each evaluation, keyphysics parameters for each event, as identified in Duke Topical Report DPC-NE-3001P-A, Multidimensional Reactor Transients and Safety Analysis Physics Parameters, (Reference Q30-1). were compared for a typical core with four MOX fuel lead assemblies and a typical all low-enriched uranium (LEU) fuel core. These physics parameters were previously submitted in Tables 3-8 and 3-10 of Attachment 3 to Reference Q30-2. These cores were of identical design with the exception that four LEU fuel assemblies in the typical all LEU fuel core were replaced with four MOX fuel assemblies in the typical MOX fuel lead assembly core. The current safety analysis values are also provided in Table Q30-1. The safety analysis values are the input values used in the Duke Power safety analysis calculations whose results are documented in Chapters 6 and 15 of the McGuire and Catawba UFSARs. The values in the safety analyses are chosen to bound expected variations in core physics parameters from cycle to cycle. The comparisons of key physics parameters for each event show no significant differences between the typical MOX fuel lead assembly core and a typical LEU core.
The variations are within the range of those expected from normal cycle-specific all LEU core designs.
Response (B)
Boron dilution during a small break loss of coolant accident (LOCA) is not part of the licensing basis for Catawba. This is a generic issue (GSI-1 85), which is currently being examined for all plants and therefore no plant specific analysis has been performed for Catawba. The MOX fuel lead assembly license amendment request shows that there are only minor changes in core reactivity parameters due to the insertion of four MOX fuel lead assemblies (See Tables 3-7 through 3-10 in Attachment 3 to Reference Q30-2).
Accordingly, the core response to a small break LOCA boron dilution event is not 59
expected to be significantly different for the MOX fuel lead assembly core than for an all LEU fuel core.
References Q30-1. DPC-NE-3001P-A, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Duke Power Company, December 2000.
Q30-2. 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.
60
Table Q30-1 Non-LOCA Transient/Accident Evaluation
,^ Translenli-dR
K; ey y yi Typical..........
Ty ic al:l LEU Currentt Safety n
Commen fl
- i-Description
Energy Release and MTC
-12.40 pcm/°F
-12.04 pcm/°F
- 0.0 pcm/°F Key physics parameters are Containment Beta-effective 0.0061 0.0062
< 0.0070 No bounded by those currently Pressure/Temperature analyzed for LEU fuel.
Response
6.2.1.4Secondary System EOC----
Key physics parameters are Pipe Ruptures and MTC
-35.92 pcm/°F
-35.61 pcm/°F 2 -51 pcm/°F No bounded by those currently Containment Pressure/
-1.64 pcm/°F
-1.64 pcm/°F s -1.20 pcm/°F analyzed for LEU fuel.
Temperature Response Beta-effective 0.0050 0.0051 s 0.0060 15.1.1 Feedwater System Bounded by excessive increase in Malfunctions that Result in NA NA NA NA No secondary steam flow analysis in a Reduction in Feedwater Section 15.1.3 Temperature 15.1.2 Feedwater System EOC----
Key physics parameters are Malfunction Causing an MTC
-35.92 pcm/°F
-35.61 pcm/°F 2 -41 pcm/°F No bounded by those currently Increase In Feedwater DTC
-1.64 pcm/°F
-1.64 pcm/°F
- -1.20 pcm/°F analyzed for LEU fuel.
Flow Beta-effective 0.0050 0.0051 2 0.0040 EOC------
15.1.3 Excessive Increase MTC
-35.92 pcm/°F
-35.61 pcm/°F 2 -51 pcm/°F No bKoeuyned by those currently in Secondary Steam Flow DTC
-1.64 pcm/°F
-1.64 pcm/°F s -1.20 pcm/°F analyzed for LEU fuel.
Beta-effective 0.0050 0.0051 2 0.0040 61
Table Q30-1 Non-LOCA Transient/Accident Evaluation TicaILE Curren~afet
-Re-
.TransientAccident:.,;
Ke~y:Physi s
'Typical MOX aa LEU'y CsrrseCntoS ents RWST boron 2 2700 ppm(1) 2 2700 ppm('"
2 2475 ppm concentration Key physics parameters are 15.1.nadert entEOC----....-
bounded by those currently Opening of a Steam Duff boron worth
-8.5 pcm/ppmt 2 e
-8.5 pcm/ppm 2 n
S -5.0 pcm/ppm No analyzed for LEU fuel.
Generator Relief or Safety Power distribution Cycle specific(3) Cycle specific(3)
Cycle specific(3)
The MTC value is modeled as a Va ve moderator density cosfficient MTC
-35.92 pcm/°F
-35.61 pcm/°F 2
-43 pcmP0F function.
-1.64 pcm/°F
-1.64 pcm/°F s -1.20 pcm/°F RWST boron 2 2700 ppm(')
2 2700 ppm(')
2 2475 ppm concentration Key physics parameters are 15.1.4 Inadvertent EOC-----
.2.
bounded by those currently 15.1.5 Stea Steam Diff boron worth
-8.5 pcm/ppm(2 ) -8.5 pcm/ppm(2
)
- -5.0 pcm/ppm No analyzed for LEU fuel.
Piping Failure Power distribution Cycle specific(3) Cycle specific(31 Cycle specific(3)
The MTC value is modeled as a Valve moderator density coefficient MTC
-35.92 pcm/°F
-35.61 pcm/°F 2 -43 pcm/°F function.
-1.64 pcm/°F
-1.64 pcm/°F
- -1.20 pcm/PF 15.2.1 Steam Pressure There are no pressure regulators in Regulator Malfunction or NA NA NA NA No the McGuire or Catawba plants Failure That Results In o
whose failure or malfunction could Decreasing Steam Flow cause a steam flow transient.
15.2.2 Loss of External NAN AN o
Bounded by turbine trip analysis in BOC-----------
Key physics parameters are 15.2.3 Turbine Trip MTC
-12.40 pcm/°F
-12.04 pcm/°F
- 0.0 pcm/°F No bounded by those currently DTC
-1.44 pcm/°F
-1.43 pcm/°F
< -0.90 pcm/°F analyzed for LEU fuel.
62
Table Q30-1 Non-LOCA Transient/Accident Evaluation Transientccident Key PhTypica MOX Typicl LEU Current Saf ety,
omments
-: s Descriptions ~.
Parameters LTA Core Value Val
.Anlysis Range 15.2.4 Inadvertent Closure Bounded by turbine trip analysis in of Main Steam Isolation NA NA NA NA No Section 15.2.3.
Valves 15.2.5 Loss of Condenser Bounded by turbine trip analysis in Vacuum and Other Events NA NA NA NA No Section 15.2.3.
Causing a Turbine Trip 15.2.6 Loss of Non-EOC---
Key physics parameters are-Emergency AC Power to MTC
-35.92 pcm/OF
-35.61 pcm/OF 2 -51 pcm/OF No bounded by those currently:
the Station Auxiliaries DTC
-1.64 pcm/0F
-1.64 pcm/0F
- -1.20 pcrm/IF analyzed for LEU fuel.
15.2.7 loss of Normal BOC Key physics parameters are 15.2.7Feedwater loss N MTC
-12.40 pcm/0F
-12.04 pcm/OF
- 0.0 pcm/0F No bounded by those currently Fpe redwateFlowDTC
-1.44 pcm/0F
-1.43 pcm/0F
- -0.90 pcm/IF analyzed for LEU fuel.
15.3.1-
~Key physics parameters are 15.2.8 Feedwater System MTC
-12.40 pcm/OF
-12.04 pcm/0F
- 0.0 pcmI0F No bounded by those currently Pipe Break DTC
-1.44 pcm/OF
-1.43 pcm/0F
- -0.90 pcm/0F analyzed for LEU fuel.
15.3.1 Partial Loss of BOC-----------
Key physics parameters are Forced Reactor Coolant MTC
-12.40 pcm/OF
-12.04 pcm/OF
- 0.0 pcm/0F No bounded by those currently Flow DTC
-1.44 pcm/0F
-1.43 pcm/OF
- -0.90 pcm/OF analyzed for LEU fuel.
15.3.2 Complete Loss of BOC-------------
Key physics parameters are Forced Reactor Coolant MTC
-12.40 pcmIOF
-12.04 pcm/OF
- 0.0 pcmI01F No bounded by those currently Flow IDTC
-1.44 pcm/01F
-1.43 pcm/01F
-0.90 pcm/OF analyzed for LEU fuel.
63
Table Q30-1 Non-LOCA Transient/Accident Evaluation
.ansient/Ac cid Key PhysiCs Typical MOX..
Typ.cal Cu..-Ty
- Rl j Cur t Safe
, S-Description -
- -,A '
Paramneters<
~ LTA CoreValue
.Core rrent Saluet analysis CommenTs
'urpShaft Seizure P ower distribution lCycle specific(3 )l Cycle specificn/ 3 CF Cycle speceific N
l Kney physic paaetr aurenl (Locked Rotor)
-1.44 pcm/°F
-1.43 pcmP0F s -0.90 pcm/°F anlyedfo LE fuel.
Bounded by reactor coolant pump 15.4 esact Clant NA NA NA NA No locked rotor analysis in Section Pump Shaft Break
~~~~~~~~~~~~~~~~~~~~~~~~15.3.3.
15.4.1 Uncontrolled Rod Ratvt diinrt 77pmsc2 77pmsc2 5pmscfrpa Cluster Control Assembly 5pcsecfr Key physics parameters are Bank Withdrawal From a si6iprcmsse No bounded by those currently Subcritical or Low Power MTC
-6.46 pcm/°F
-6.18 pcml°F
- 6
.0 pcm/°e F
analyzed for LEU fuel.
Startup Condition DTC
-1.69 pcm/°F
-1.67 pcm/°F 7-0.90 pcm/lF Ex-core detector signal 0.7% (2) 0.7% (2)
Max change @ 50%
(indicated power) power
- 5%.
11.3% (2) 11.3% (2)
Max change @ 10%
()
power s 12%.
Reactivity addition rate 18 pcm/sec(2) 1827 pcm/seC
- 45 pcm/sec 15.4.2 Uncontrolled Rodpressome Key physics parameters are Cluster Control Assembly BOC---
r No bounded by those currently Bank Withdrawal at Power MTC
-12.40 pcm/°F
-12.04 pcm/IF
- 0.0 pcm/°F analyzed for LEU fuel.
-1.44 pcm/°F
-1.43 pcm/IF s -0.90 pcm/iF MTC
-35.92 pcm/°F
-35.61 pcm/°F s -24 pcm/°F RDTC
-1.64 pcm/lF
-1.64 pcm/F S -1.20 pcml°F 64
Table Q30-1 Non-LOCA Transient/Accident Evaluation TransientAccldent Key PhysIcs Typical MOX
-Typical LEU Current Safety analysis Coments'-
-' Description' Parameters';
LTA Core Value,,Core Value:
- Analysis Range BOCIMOCIEOC-Dropped rod worth 615 pcm(2) 615 pcm(2) s 700 pcm (small to max)
Excore detector tilt N/A(5)
N/A(
5
)
NW')
(Min indicated power)
Power distribution Cycle specific(3) Cycle specific(3)
Cycle specific(3 Peaking)
BOC ---- ---
MTC
-12.40 pcm/0F
-12.04 pcm/°F
- -7.0 pcm/°F 15.4.3 Rod Cluster Control DTC
-1.44 pcmn/ 0F
-1.43 pcrnI0F
- -0.90 pcmn/0F Kypyisprmtr r
Assembly Misoperation Available rod worth for 31.0 pcm (2 )
310 pcm(2)
- 350 pcm No bounded by those currently (System Malfunction or withdrawal analyzed for LEU fuel.
Operator Error) - Rod Drop MOC lze orLU ul MTC
-16.85 pcm/°F
-16.47 pcm/0F S -10.0 pcm/0F DTC
-1.49 pcm°IF
-1.48 pcm/°F s -1.05'pcm/PF vailable rod worth for 319 pcm(2) 319 pcm(2)
- 450 pcm withdrawal EOC ---- --- h--
.MTC DTC
-35.92 pcm/°F
-35.61 pcm/0F s -24.0 pcm/°F itable rod worth for
-1.64 pcm/ 0F
-1.64 pcm/ 0IF
- -1.20 pcm/0F withdrawal 398 pcm (2) 398 pcm (2)
- 500 pcm 15.4.3 Rod Cluster BOC---------
2 Control Assembly Single rod worth 82 pcmt2) 82 pcm2 s 105 pcm Misoperation (System Power distribution Cycle specifict3) Cycle specific(3 )
Cycle specific 3 N
Key physics parameters are Malfunction or Operator Ex-core tilt 0.95(2) 0.95(2)
> 0.9 No bounded by those currently Error) - Single Rod MTC
-6.46 pcm/IF
-6.18 pcm/0F 7.0 analyzed for LEU fuel.
Withdrawal DTC
-1.69 pcm/°F
-1.67 pcm/0F s -0.90 pcm/°F 65
Table Q30-1 Non-LOCA Transient/Accident Evaluation
-.:T r
ansidnV
~
ccid ent
~
.:Key hysics.-
- Typical MOX-: ',TypicalLEU -
CurrentSafety.
R
-ransientlAccien Kebhsc-aayi omnme~nts
--SDescriptio
., Parameter. ;, '
L TA reVal e
- h.
Ce Valu Anaysis Range Re lCed?
15.4.4 Startup of an EOC ------
Key physics parameters are Inactive Reactor Coolant MTC
-35.92 pcm/0F -35.61 pcm/0F
> -51 pcm/0F No bounded by those currently Pump at an Incorrect DTC
-1.64 pcm/OF
-1.64 pcm/OF
- -1.20 pcm/OF analyzed for LEU fuel.
Temperature 15.4.6 Chemical and Initial boron Volume Control System concentration - closest Cycle specific Cycle specific Cycle specific Malfunction that Results in to critical a Decrease in Boron t cat boron Concentration in the oncenbron Cycle specific Cycle specific Cycle specific Reactor Coolant ihs 15.4.7 Inadvertent Loading and Operation of a Fuel NA NANA NA No Note (6)
Assembly in an Improper Position 66
Table Q30-1 Non-LOCA Transient/Accident Evaluation
..tTran:siVAntiAccidnt, Ke Physics.-
Tyica iox Tyica 1EV CurentSafety Re-a Description; Parmeters LTA Core Value CoreValue
" -Analysis Rang R
i C
HZP ejected rod worth 349 pcm 375 pcm
- 720 pcm HFP ejected rod worth 43 pcm 44 pcm
- 200 pcm HZP total peak. factor 9.5(2) 93 '(2)
- 9.8 HFP total peak. factor 3.1 (2) 3.1 (2)
- 4.32 HZP MTC
-8.17 pcm/°F
-7.82 pcm/lF s 7.0 pcm/°F HFP MTC
-13.12 pcm/°F
-12.71 pcm/°F
- 0.0 pcm/0F HZP DTC
-1.74 pcm/°F
-1.73 pcm/°F s -0.9 pcm/°F HFP DTC
-1.41 pcm/PF
-1.40 pcm/°F s -0.9 pcm/°F 15.4.8 Spectrum of Rod HZP Beta-effective 0.0061 0.0062 2 0.0055 (7)
See responses to Cluster Control Assembly Yes Reactor Systems Ejection Accidents EOC ------------------
Questions 32 and 33 HZP ejected rod worth 412 pcm 406 pcm s 900 pcm HFP ejected rod worth 47 pcm 46 pcm s 200 pcm HZP total peak. factor 13.3(2) 13.3(2) s 15.46 HFP total peak. factor 3.0(2) 3.0(2) s 4.39 HZP MTC
-25.55 pcm/°F
-25.32 pcm/°F s -10.0 pcm/°F HFP MTC
-37.48 pcmP0F
-37.18 pcm/°F s -24.0 pcm/°F HZP DTC
-1.88 pcm/°F
-1.87 pcm/ 0F s -1.2 pcm/°F HFP DTC
-1.51 pcm/°F
-1.51 pcm/°F s -1.2 pcm/°F Beta-effective 0.0050 0.0051 2 0.0040 15.5.1 Inadvertent All thermal margins increase during Operation of Emergency NA NA NA NA No this transient. There are no Core Cooling System important physics parameters.
During Power Operation 67 0
Table Q30-1 Non-LOCA Transient/Accident Evaluation Tra sien/A cident -
e Physics I Ty. pical MOX Typical LEU'.
Current Safety
-. ;me'tS
'.D'escription'.
.'.-.- > Paramneterso'
-TA Core Value Core Value
-;Analysis'Range R
?i -
C-15.5.2 Chemical and Bounded by inadvertent operation Volume Control System NA NANA NA No of emergency core cooling system Malfunction That Increases during power operation analysis in Reactor Coolant Inventory Section 15.5.1.
15.6.1 Inadvertent There are no important physics Opening of a Pressurizer NA NA NA NA No parameters associated with this Safety or Relief Valve event.
15.6.2 Break In Instrument Line or Other Lines From There are no important physics Reactor Coolant Pressure NA NA NA NA No parameters associated with this Boundary That Penetrate event.
Containment There are no important physics 15.6.3 Steam Generator NA NA NA NA No parameters associated with this event.
68
Table Q30-1 Non-LOCA Transient/Accident Evaluation Notes:
(1) This is the required minimum RWST concentration based on the assumed safety analysis value. This value will not change for the MOX fuel lead assembly core.
(2) This value is not explicitly calculated for a MOX fuel lead assembly core but is taken from a typical all LEU fuel core (McGuire 1 Cycle 16).
The value will change within normal variations based on cycle design changes, but the MOX fuel lead assembly value will remain within the analyzed value. This will be confirmed when the actual MOX fuel lead assembly core design is finalized.
(3) Relative to a typical LEU core design, the core power distribution Is expected to be similar for the MOX fuel lead assembly core. It is evaluated for each cycle and will be evaluated for the actual MOX fuel lead assembly core design when finalized.
(4) Relative to a typical LEU core design, the initial and critical boron concentrations are expected to be similar for the MOX fuel lead assembly core design when finalized.
(5) Ex-core tilt for the rod drop analysis Is analyzed generically as described in DPC-NE-3001 P-A, Multi-Dimensional Reactor Transients and Safety Analysis Physics Parameters Methodology.
(6) See the discussion on misloaded fuel assembly in Section 3.7.2.4 of Attachment 3 of Duke's license amendment request dated February 27, 2003.
(7) Re-analysis performed due to MOX fuel-specific callgm acceptance criterion.
Acronyms BOC beginning of cycle DTC Doppler temperature coefficient EOC end of cycle HFP hot full power HZP hot zero power LOCA loss of coolant accident MOC middle of cycle MTC moderator temperature coefficient RWST refueling water storage tank 69
- 31. Section 3.7.2.2 discusses the thermal-hydraulic differences for the different co-resident fuel types. Please provide the limits analyzed for the different fuel types, including a discussion on how they were obtained.
Response (Previously submitted October 3, 2003)
The co-resident fuel types, Westinghouse RFA and Mark Bw/MOX1, will be analyzed with their respective critical heat flux (CHF) correlations and limits. The RFA fuel will be analyzed with the Westinghouse WRB-2M CHF correlation with a design limit deparature from nucleate boiling ratio (DNBR) value (95/95) of 1.14 for deterministic analyses and a design limit DNBR value of 1.30 for analyses where uncertainties are combined statistically.
The Mark BW/MOX1 fuel will be analyzed with the Framatome BWU-Z CHF correlation with a design limit DNBR value (95/95) of 1.19 for deterministic analyses and a design limit DNBR value of 1.36 for analyses where uncertainties are combined statistically.
Derivation of these limits is described in References Q31-1, Q31-2, and Q31-2.
References Q31-1. DPC-NE-2005P-A, Revision 3 (Appendix E), Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology, September 2001.
Q31-2. DPC-NE-2009P-A, Revision 2, Duke Power Company Westinghouse Fuel Transition Report, December 2002.
Q31-3. WCAP-1 5025P, Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 1 7x] 7 Rod Bundles with Modified LPD Mixing Vane Grids, February 1998.
70
- 32. On page 3-31, the first paragraph states that no MOX assembly will be rodded. Will that always be true for both LTAs and batch loading? Also, is there a scenario where the MOX assembly rod ejection energy deposition will be higher than that for an LEU?
Please provide the worth of the most reactive rod for an all-LEU core and for a core with four MOX assemblies. Also, provide a discussion on what are considered to be the appropriate conservatisms for the ejected rod analysis.
Response
The first paragraph of page 3-31 of the February 27, 2003 MOX fuel lead assembly license amendment request (Reference Q32-1) states that "in the first cycle of operation MOX fuel assemblies with be unrodded." The MOX fuel lead assembly license amendment request deals only with operation with MOX fuel lead assemblies. It is contemplated that in batch use MOX fuel may be operated in rodded core locations.
The lead assembly cores will be designed such that the peak energy deposition (calgm) resulting from the worst-case control rod ejection will be in LEU fuel, not in a MOX fuel lead assembly. For representative cores containing (i) all LEU fuel, and (ii) 189 LEU fuel assemblies and four fresh MOX fuel assemblies, Table Q32-1 shows the nominal hot zero power (HZP) worth of the most reactive control rod that is available for ejection from the core. These results show that the addition of four MOX fuel lead assemblies has no significant impact on the worth of the most reactive control rod.
In the end of cycle (EOC) HZP control rod ejection simulation for a representative core containing four MOX fuel lead assemblies, conservative values were used for Peff, ejected rod worth, fuel temperature coefficient, and moderator temperature coefficient. The nominal values and the conservative values assumed for the analysis are listed in the response to Reactor Systems Question 33. Parf is reduced by 20%, ejected rod worth is increased by 55%, the magnitude of the fuel temperature coefficient is decreased by 36%,
and the magnitude of the moderator temperature coefficient is decreased by 60%. These large adjustments are made to the key parameters to demonstrate that the peak fuel enthalpy response is benign even with very large conservatisms applied.
In addition, conservative assumptions were made for control rod trip worth and trip delay time. The control rod trip worth was reduced by approximately 50% relative to the nominal value. The reactor was assumed to trip following the receipt of a high power signal from three of the four excore detectors. A conservative trip delay time of 2.7 seconds was assumed between the generation of the trip signal and beginning to take credit for control rod movement into the core. The peak energy deposition is not sensitive to these control rod trip parameters, because the fuel temperature coefficient turns around the transient well before the control rods enter the core.
The conservatisms used for these MOX fuel lead assembly analyses should not be taken as setting a precedent for future analyses of control rod ejection analyses in partial MOX fuel cores. For batch use of MOX fuel, Duke intends to submit a safety analysis methodology topical report to the NRC in 2004.
71
Refe-.ence Q32-1. 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 Table Q32-1 Rod Worth vs Burnup Number of MOX Nominal Worth of Most Reactive Control Rod Assemblies in Core Available for Ejection from the Core (pcm)
349:
412 0
375 406
- 33. On page 3-31, the second paragraph states that the rod ejection accident was conducted at the end-of-cycle hot zero power condition. Is this supposed to indicate that this condition provides the most limiting event? If so, provide an explanation on how and why it is the most limiting condition for the rod ejection accident with four MOX LTAs in the core.
Response
From the standpoint of change in fuel enthalpy, the most limiting case is typically end of cycle (EOC), hot zero power (HZP) conditions. Analysis at HZP conditions allows for greater positive reactivity insertion from the ejected rod prior to temperature feedback stopping the neutron power increase. EOC is characterized by higher worth ejected rods and lower effective delayed neutron fraction (Per), which make the results worse.
However, EOC is also characterized by more negative fuel temperature and moderator temperature coefficients, which make the transient response less severe. It should also be noted that the introduction of conservatism to key parameters such as those discussed above can affect the determination of which condition is most limiting.
Control rod ejection analyses were performed using the SIMULATE-3K MOX computer code for HZP and hot full power (HFP) conditions at both beginning and end of cycle in the representative core containing four fresh MOX fuel lead assemblies. The input assumptions and results are summarized in Table Q33-1.
Transient simulations using nominal and conservative nuclear parameter inputs were run.
All of the nominal case analyses had MOX fuel pellet enthalpy increases of less than 1 cal/gm. This was due to the low (much less than $1) reactivity insertions from the ejected rods.
72
Table Q33-1 Inputs and Results for Core Control Rod Ejection Analyses in a Representative MOX Fuel Lead Assembly Core
';'Parameter BOC BOC EOC~ t:EOC'
- __Parameter__-__. ___- __-_.
HZP HFP HZP. -HFP LEU Fuel Peak Pellet Enthalpy (callgm) (See note)
Initial Steady-State 17.1 75.7 17.1 76.0 Maximum 50.4 77.7 54.1 76.8 MOX Fuel Peak Pellet Enthalpy (callgm) (See note)
Initial Steady-State 17.1 72.1 17.1 65.8 Maximum 36.5 73.6 31.6 66.5 Key Parameter Best Estimate Value Core Average Beta Effective 0.00608 0.00609 0.00504 0.00504 Ejected Rod Worth (pcm) 349 43 412 47 Fuel Temperature Coefficient (pcm/0F)
-1.74
-1.41
-1.88
-1.51 Moderator Temperature Coefficient (pcm/IF)
-8.17
-13.12
-25.55
-37.48 Key Parameter Conservative Value Core Average Beta Effective 0.00481 0.00482 0.00399 0.00399 Ejected Rod Worth (pcm) 543 67 641 73 Fuel Temperature Coefficient (pcm/OF)
-1.10
-0.89
-1.19
-0.96 Moderator Temperature Coefficient (pcm/OF)
-3.19
-5.12
-9.98
-14.64 Note:
Cal/gm values in this table are taken directly from SIMULATE-3K MOX. These include conservative nuclear input assumptions but do not include thermal-hydraulic conservatisms (see text of response).
74
- 34. Section 3.7.3 discusses the radiological consequences of selected postulated accidents where MOX fuel lead assemblies have the potential to affect the dose consequences.
Please explain how the source term used for these analyses was developed?
Response
The MOX fuel source term was developed utilizing the Standard Computer Analyses for Licensing Evaluation (SCALE, References Q34-1 and Q34-2) computer code, version 4.4, from the Oak Ridge National Laboratory (ORNL) code collection in the Radiation Safety Information Computational Center (RSICC). It was chosen because of its flexibility, its ability to produce problem specific libraries, and its ability to update these libraries during the code's execution of the specific problem. This provides for more refined and more accurate results when compared with codes which utilize "pre-packaged" libraries based on parameters that may not precisely match those of the specific problem. In addition, this flexibility provides more accurate modeling of MOX fuel since SCALE will include those isotopes and reactions which may not be included in the LEU based "pre-packaged" libraries of other codes (i.e. ORIGEN2). Duke is experienced with source term generation using SCALE.
SCALE is comprised of modules which support a variety of purposes. For source term development, the SAS2H module (Reference Q34-1) is used. This module is commonly used for the development of source terms by modeling fuel assembly depletion and the decay of resulting fission products. An internal 44 energy group neutron cross section library collapsed from a 238 group library based upon the ENDF/B-V data was used as the basis for the depletion models and calculations. The code and its data and cross section libraries have received published validation for LEU and MOX fuel in the United States (mainly ORNL) and in Europe (References Q34-3 through Q34-6).
SAS2H uses the one dimensional discrete ordinates code XSDRNPM for neutronics modeling and the ORIGEN-S point depletion code for depletion and transmutation calculations. The computations are performed using multiple passes in a "two path" scheme whereby the first path (Path A) represents the fuel by a pin-cell lattice and generates the cell-weighted (homogenized) fuel region cross sections. Path B utilizes a multi-region assembly model where fuel cross sections and flux are used to update the ORIGEN-S nuclear data. Spent fuel isotopics are produced in the last ORIGEN-S pass.
Many cases were calculated to arrive at a spectrum of MOX fuel source predictions which included various combinations of concentration, burnup, time in core life, and peaking factors for each general accident scenario. The isotopic inventories in these output files can be extracted and processed to suit the specific needs of the analysis.
See the response to Radiological Consequences Question 3(f) for further information related to these specific applications.
75
References Q34-1. SCALE Version 4.4, A Modular Code System for Performing Standardized Compiuter Analyses for Licensing Evaluation, Oak Ridge National Laboratory, ORNIJNUREG/CSD-2/R6, May 2000.
Q34-2. SCALE 4.4, A Modular Code System for Performing Standardized Comnputer Analyses for Licensing Evaluation for Workstations and Personal Compluters, Volume 0, Oak Ridge National Laboratory, April 1999.
Q34-3. Validation of the SCALE System for PWYR Spent Fuel Isotopic Composition Analysis, ORNL/TM-12667, Revision 0, March 1995.
Q34-4. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictionsfor PWVR Spent Fuel, ORNLUTM-13317, Revision 0, September 1996.
Q34-5. SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volulme 2 - Sequoyah Unit 2 Cycle 3, Bowman, Hermann and Brady, ORNL/TM-12294tV2, March 1995.
Q34-6. Benchmark of SCALE (SAS2H) Isotopic Predictions of Depletion Analyses for San Onofre PWR MOXFzuel, ORNIJTM-1999/326, Herman, Feb 2000.
- 35. Provide information on how the peaking factor limits used for MOX fuel were developed.
Response (Previously submitted October 3, 2003)
During the August 13, 2003 telephone conference among Duke, Framatome ANP, and the NRC staff, the staff clarified that the specific peaking factor limits of interest are LOCA peaking limits as provided in Table 3-6 of the February 27, 2003 License Amendment Request. This response was developed accordingly.
The MOX fuel assemblies are limited to operation within the allowed peaking presented in Table 3-6 of the license amendment request. The allowed peaking was established using the same approach as for any fuel design. A target allowable peaking (Limiting Condition for Operation) was determined considering plant operability and the required fuel service.
The acceptability of that target peaking was then confirmed through appropriate analysis (design, safety, and LOCA). The considerations included in establishing the MOX fuel assembly target limit for normal operation were:
a) MOX fuel lead assemblies are demonstration lead assemblies and it is therefore appropriate to control them to somewhat reduced limits b) Recognition that demonstration lead assemblies should be operated with fairly high duty in order to promote the discovery of issues that should be addressed prior to batch implementation, c) The insertion of four demonstration lead assemblies should not impose an operational hardship on the irradiating plant.
As discussed in Section 3.7.1.4 of Reference Q35-1, additional LOCA studies were performed after the submittal of the license amendment request to further refine the LOCA peaking limits for the MOX fuel lead assemblies. These studies included:
76
a) Time-in-life (burnup) sensitivity to determine any burnup dependent limitation on the fuel, and b) Axial power distribution sensitivity to confirm the Kz curve.
The results of these studies are discussed in the response to Reactor Systems RAI Question 14.
Reference Q35-1. 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.
- 36. Attachment 3-1 contains the criticality evaluation for MOX fuel storage in the spent fuel pools. For each of the spent fuel pool criticality analyses, provide the appropriate regulatory criteria used for the analysis and show how the criteria are met.
Response
Page A3-8 in Appendix 3-1 of Attachment 3 to Reference Q36-1 notes that 10 CFR 50.68 (b) is the governing regulation for the spent fuel pool (SFP) criticality evaluations. In accordance with 10 CFR 50.68 (b) (4), the Catawba spent fuel pool criticality analysis for normal conditions, which does not take credit for soluble boron, must satisfy the following criterion:
The equation defining the maximum 95/95 kdrs is presented on page A3-9 of Appendix 3-
- 1. For normal conditions, the Catawba criticality analysis yielded a maximum 95/95 kff of 0.9217 with no soluble boron in the SFP. For the worst-case accident condition (weir gate drop), the maximum 95/95 kr was 0.9429, with full credit for soluble boron (2700 ppm) in the Catawba SFP per the double contingency principle. Numerical results for the individual bias and uncertainty components in these maximum 95/95 kefT values are presented in the response to Reactor Systems Question 39.
Reference Q36-1. Tuckman, M. S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendment 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.
- 37. Table A3-1 on page A3-3 lists the minimum required boron concentration in each of the spent fuel pools. Is this concentration required to maintain the k-effective < 0.95, or is this the concentration in the refueling water storage tank?
77
Response
The minimum SFP boron concentrations listed in Table A3-1 are the minimum values currently specified in the Core Operating Limits Report (COLR) for Catawba. While the current COLR for each Catawba unit specifies the same minimum boron concentration for both the spent fuel pool and the Refueling Water Storage Tank, the Catawba spent fuel pool criticality analysis confirms Keff < 0.95 at normal conditions with no credit for soluble boron.
- 38. On page A3-6, reference is made to bench-marking calculations by Duke. (A) What were the highest MOX and LEU enrichments assumed in the calculations? (B) Was the MOX weapons-grade MOX?
Response (A)
For the benchmark MOX fuel critical experiments evaluated in support of this license amendment request, the highest plutonium concentration analyzed was 19.7 weight percent (w/o) total plutonium (17.3 w/o fissile plutonium) in the MIX-COMP-THERM-001 experiments performed by Battelle - see Table A3-3 in Appendix 3-1 of Attachment 3 to Reference Q38-1. All MOX fuel critical experiments contained natural uranium.
Previously benchmarked LEU critical experiments evaluated U-235 enrichments as high as 4.31 w/o. As noted in Appendix 3-1, the nominal Mark-BW/MOX1 lead assembly has a fissile plutonium concentration of 4.15 w/o with depleted uranium at 0.25 w/o U-235.
The LEU filler fuel assemblies evaluated for MOX/LEU fuei storage configurations in the Catawba spent fuel pools have fissile concentrations below 2.0 w/o.
Response (B)
The Department of Energy classifications of plutonium are as follows: weapons grade (WG) is 240Pu < 7.0%, fuel grade is 240Pu between 7% and 20%, and reactor grade (RG) is 24OPu > 20%. The benchmark MOX fuel critical experiments had 240Pu concentrations ranging from 7.8 to 22.0 wt. % in Pu, indicating that it ranges from near WG to good quality RG. Investigation of benchmark results does not indicate a trend in the method bias or uncertainty over the range of 240Pu concentration. Furthermore, as discussed in Section 3.1 of Reference Q38-2, it is reasonable to expect that nuclear analysis methods that are capable of analyzing RG MOX fuel and LEU fuel will perform with comparable accuracy for WG MOX fuel. Therefore, it is judged acceptable to apply the benchmark bias and results to the Catawba spent fuel pool storage analysis for WG Mark-BW/MOXl lead assemblies.
Reference Q38-1. 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.
Q38-2. BAW-10238(NP) Revision 1, MOXFuel Design Report, Framatome ANP, May 2003.
78
- 39. On page A3-7, reference is made to the method bias and uncertainty with respect to the MOX fuel critical experiments. Is this the KENO model calculational bias or the uncertainty associated with the experiments? Please clarify what the bias and uncertainty are related to. In addition, for each spent fuel pool criticality calculation, provide, in a tabulated form, a list of all the uncertainties accounted for in each of the k-effective calculations. Please provide the information for the cases with and without boron. In each case, please state the regulatory criteria associated with spent fuel criticality that needs to be satisfied, and how this criteria has been met.
Response
The method bias and method uncertainty discussed and quantified on page A3-7 are measures of how well SCALE 4.4 / KENO V.a models the selected MOX critical experiments. These are more fully described as the benchmark method bias and benchmark method uncertainty on pages A3-l0 and A3-1 1.
For the Catawba SFP criticality analysis, which analyzed MOX and LEU fuel stored together in accordance with the first configuration shown in Figure A3-5, Table Q39-1 below shows the results of the maximum 95/95 kfr calculation for normal (non-accident) conditions, including all the individual bias and uncertainty contributors. A bounding benchmark method bias of +0.0045 Ak was used, based on the trends observed in the MOX fuel critical experiment benchmarks with respect to average fission energy. Note that, because the Catawba spent fuel pool storage racks do not contain any poison panels, the Boraflex uncertainties listed in Table A3-4 of Appendix 3-1 do not apply. Note also that in addition to the other biases and uncertainties listed in Table A3-4, Table 1 provides conservative reactivity uncertainties for fuel rod cladding diameter, assembly rack positioning, and storage cell ID. These uncertainties were included in the criticality analysis to ensure overall conservatism of the computational results, and are described below:
Fuel Rod Cladding Diameter Manufacturing Uncertainty A change in the cladding outer diameter (OD) affects the amount of moderation within the fuel assembly lattice. The MOX fuel assembly lattice is undermoderated, so a reduction in cladding OD should slightly increase the MOX fuel reactivity. The tolerance on the cladding diameter is + 0.002 inches. This tolerance is similar to that for current LEU fuel cladding, and was also applied to the LEU filler fuel used in the Catawba storage configurations.
Assembly Rack Position Uncertainty This uncertainty accounts for the fact that an assembly in a storage cell may not be perfectly centered within that cell, but closer to one or two walls of that cell. The rack position uncertainty model takes a cluster of four storage cells and pushes the assemblies within those cells as close as possible to each other or as far from each other as possible.
The positioning that produces the largest reactivity increase over the nominal (centered) positions is used as the uncertainty contributor.
79
Storage Cell Inside Dimension (ID) Uncertainty This uncertainty accounts for possible variations in the rack storage cell inner distance between cell walls. For conservatism, the storage cell pitch was also allowed to decrease in the case of a cell ID reduction, making that the bounding case. Drawings for the Catawba spent fuel pool storage racks show a + 0.063-inch tolerance on storage cell ID.
As noted in the response to Reactor Systems Question 36, the sole nornal-condition sub-criticality criterion for Catawba spent fuel pool storage is that the maximum 95/95 keff remain below 0.95 with no boron in the water. The results in Table Q39-1 demonstrate that this criterion has been met.
Table Q39-2 provides the results of the maximum 95/95 kff calculation for the worst-case accident condition (weir gate drop) in the Catawba spent fuel pool, including all the individual bias and uncertainty contributors. Note that there are no uncertainties for storage rack center-to-center cell spacing, assembly rack position, or storage cell ID associated with the weir gate drop accident, because the model for this accident considers the fuel and storage racks in an optimum-reactivity arrangement. Per the double contingency principle, full credit for the minimum soluble boron in the Catawba spent fuel pool (2700 ppm) may be taken for accident conditions. As Table Q39-2 shows, the maximum 95/95 kfr for accident conditions in the Catawba spent fuel pool remains below 0.95.
Table Q39-1 Maximum 95/95 kff Calculation for Normal Conditions for MOX Fuel Lead Assembly Spent Fuel Pool Storage (0 ppm Boron in 68 IF - 150 OF Water)
-: : Parameter Value,:
Nominal keff 0.9028 Benchmark Method Bias 0.0045 Nominal keff uncertainty 0.0030 Benchmark Method Uncertainty 0.0075 Plutonium Concentration Manufacturing Uncertainty 0.0033 Fuel Density Manufacturing Uncertainty 0.0030 Storage Rack Cell Wall Thickness uncertainty 0.0032 Storage Rack Center-to-Center Cell Spacing 0.0044 Uncertainty Fuel Rod Cladding Diameter Uncertainty 0.0032 Assembly Rack Position Uncertainty 0.0083 Storage Cell ID Uncertainty 0.0035 Maximum no-Boron 95/95 kff 0.9217 80
Table Q39-2 Maximum 95/95 keff Calculation for Weir Gate Drop Accident for MOX Fuel Lead Assembly Spent Fuel Pool Storage (2700 ppm Boron in 68 OF - 150 OF Water)
Parameter '
Value Nominal k ff 0.9260 Benchmark Method Bias 0.0045 Nominal kff uncertainty 0.0030 Benchmark Method Uncertainty 0.0075 Plutonium Concentration Manufacturing Uncertainty 0.0055 Fuel Density Manufacturing Uncertainty 0.0058 Storage Rack Cell Wall Thickness uncertainty 0.0035 Storage Rack Center-to-Center Cell Spacing Uncertainty Fuel Rod Cladding Diameter Uncertainty 0.0036 Assembly Rack Position Uncertainty Storage Cell ID Uncertainty Maximum Accident-Condition 95/95 kef 0.9429 81
- 40. Page A3-8 discusses the spent fuel pool racks in the third paragraph. Please state if MOX fuel (new and spent) will be stored in racks specifically designed for MOX fuel or will they be stored in regular storage racks.
Response
The MOX fuel lead assemblies (fresh and irradiated) will be stored in the existing (regular) Catawba spent fuel pool storage racks.
- 41. Page A3-8 frequently references a nominal model. Please specify all of the parameters for the nominal model, including enrichment, size, etc.
Response
The nominal model refers to the nominal specifications for the Mark-BW/MOXl fuel assembly, the highest-reactivity LEU fuel assembly, and the spent fuel pool storage racks.
The MOX and LEU fuel assembly model data are provided in Table 3-1 of Attachment 3 of Reference Q41-1. For the SCALE 4.4 / KENO V.a model of the MOX and LEU fuel assemblies, average fuel stack densities of 10.32 grams/cc and 10.34 grams/cc for MOX fuel and LEU fuel, respectively, were derived from the data in Table 3-1. Spent fuel pool storage rack information (including thickness and pitch for the stainless steel rack cells) is shown in Table A3-1 in Appendix 3-1. The isotopic composition for the nominal MOX fuel lead assembly is given in Table A3-2 in Appendix 3-1. As noted on page A3-12, the criticality analysis conservatively maximized the fissile plutonium content of the Table A3-2 data.
Reference Q41-1. Tuckman, M.S., February 27, 2003 Letter to U.S. Nuclear Regulatory Commission, Proposed Amendment 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.
- 42. On page A3-9, the second and third paragraphs make reference to criticality calculations and boron concentrations for various analyses without specifying actual values. Provide the actual values of k-effective calculated (including all the uncertainties) for each case, and the respective boron concentration needed to meet the applicable k-effective value.
Provide an example of one of the calculations performed.
Response
The response to Reactor Systems Question 39 provides detailed results, including all biases and uncertainties, for the maximum 95/95 keff calculations (normal and accident conditions) in the Catawba spent fuel pools.
82
- 43. Provide the reference for the plutonium concentration uncertainty value used on page A3-11.
Response
The value for the plutonium concentration manufacturing uncertainty (+/- 0.075 w/o plutonium) was provided by Framatome ANP. They stated that a lot tolerance of 4 1.5%
could be applied to the MOX fuel plutonium concentration. Figure A3-6 in Appendix 3-1 shows that the highest total plutonium concentration for an individual fuel pin in the Mark-BW/ MOXI fuel assembly is 4.94 w/o. If this is conservatively increased to 5.0 w/o the tolerance on the plutonium concentration is (+/- 1.5 %) x (5.0) = 4 0.075 w/o plutonium.
- 44. Page A3-1 1 contains a discussion of the fuel density manufacturing uncertainty. Was the maximum tolerance used in the calculation for maximizing the fuel density?
Response
The maximum tolerance values for the manufactured fuel density were used. Three parameters are integral to the determination of the manufactured fuel density: pellet diameter, pellet densification and pellet dishing/chamfer factor. Of these parameters, the pellet diameter and pellet densification use the maximum tolerance values for the Mark-BW/MOXl fuel assembly design (i.e., [
] increase in pellet diameter, and
+1.5% in pellet densification). The dish/chamfer factor is not specified for the MOX fuel, therefore, the maximum tolerance for a mechanically equivalent LEU fuel assembly design was used (i.e., 0.4% reduction of dish/chamfer factor). For further conservatism, the manufactured fuel density uncertainty was calculated for the simultaneous upper tolerance of these three parameters, which is conservative, because in reality they are statistically independent.
The maximum tolerance values for the pellet dishing reduction (from a [
3 dishing/chamfer factor to [
]), pellet diameter increase (from 0.3225 to[
]
inches), and pellet densification increase (from 95 % to 96.5 % theoretical density) were used to maximize the MOX fuel density. The nominal density of 10.32 grams/cc for MOX fuel (as noted in the response to Reactor Systems Question 41) is thus maximized:
(10.32 g/cc) x (0.9940/0.9893) x (0.965/0.95) x (0.3230/0.3225)2 = 10.565 g/cc Provide the technical basis for the fuel density manufacturing, storage rack cell wall thickness, and storage rack center-to-center cell spacing tolerances referred to on page A3-11.
Response
The tolerance for the manufactured fuel density is described in the response to Reactor Systems Question 44. The values are based on fuel design documents from Framatome ANP.
The tolerance for the storage rack cell wall thickness manufacturing uncertainty (+/- 0.01 inches) was based on an estimation of the stainless steel thickness tolerance for the 83
Catawba spent fuel po6l storage cells using standard thickness tolerance data for steel plates.
The tolerance used for the storage rack center-to-center cell spacing uncertainty (+/- 0.125 inches for the Catawba spent fuel pool storage racks) was taken from vendor drawings.
The storage rack tolerances are the same values as have been used for previous Catawba LEU fuel storage criticality analyses.
- 46. Pages A3-9 thru A3-12 present the 95/95 calculation that was performed for the spent fuel pool criticality analysis. Please clarify whether the 95/95 calculation presented is the upper bound calculation.
Response
Yes, it is an upper bound calculation. The maximum 95/95 kff calculations discussed in Appendix 3-1 and tabulated in the response to Reactor Systems Question 39 were performed in accordance with the guidance in ANSI/ANS-57.2-1983 and in an August 19, 1998 NRC memorandum from L. Kopp to T. Collins. Each maximum 95/95 kfr determined in this manner is an upper bound value in that there is at least 95 percent confidence that the actual kef has a 95 percent probability of being less than the calculated 95/95 kfr.
- 47. Attachment 6 contains the justification for exemptions to the regulations to use M5 cladding and MOX fuel. As stated in the attachment, exemptions are allowed under Title 10 of the Code of Federal Regulations 10 CFR 50.12 (a)(2)(ii) if the proposed change meets the underlying purpose of the regulation. Therefore, please provide a more detailed discussion of the underlying purpose of the individual requirements of the regulation (i.e. Baker-Just equation) that apply to the exemption and how the proposed change meets the underlying purpose of that individual requirement of the regulation.
Response
The underlying purpose of 10CFR 50.46 is to ensure that light water reactors have an adequate emergency core cooling system (ECCS) to mitigate a postulated LOCA. To assess the adequacy of the ECCS design, 10CFR 50.46 establishes five acceptance criteria.
These are listed in Table Q12-1 along with their applicability to MOX fuel.
With regard to the exemption request for M5Tm cladding use, paragraph I.A.5 of Appendix K requires that the Baker-Just equation be used in the ECCS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation. The Baker-Just equation is known to provide a conservative representation of Zircaloy cladding oxidation. To verify that the Baker-Just equation is similarly appropriate for application to M5Tm cladding, Framatome ANP conducted high temperature oxidation tests. At high temperatures the oxidation rates for M5T1M alloy and Zircaloy-4 are essentially the same. At lower temperatures, the M5`r' oxidation (corrosion) rate is substantially lower than Zircaloy-4. For both cladding materials, the Baker-Just equation conservatively bounds the data. This information is documented in Reference Q47-1.
84
Therefore, the required cladding oxidation model (Baker-Just equation) is appropriate for application to M5 advanced alloy cladding material and strict application of the portion of I OCFR 50, Appendix K which refers solely to Zircaloy and ZIRLO as cladding materials is not necessary to achieve the underlying purpose of I OCFR 50.46.
The underlying purpose of 1 OCFR 50.44 is to control the amount of hydrogen gas produced following a postulated loss of coolant accident from 1) metal-water reaction involving the fuel cladding and the reactor coolant, 2) radiolytic decomposition of the reactor coolant, and 3) corrosion of metals. The only hydrogen production source affected by the use of M5Th1 cladding is the metal-water reaction of the cladding with the reactor coolant. As discussed in the previous paragraph, the oxidation rate of M5TM cladding is lower than that of Zircaloy-4 so that with the granting of the requested exemption, the underlying purpose of the regulation would still be satisfied.
Reference Q47-1. BAW-1 0227P-A, Evaluation of Advanced Cladding and Stnictural Material (MSTAI) in PWR Reactor Fuel, February 2000.
85
RADIOLOGICAL CONSEQUENCES QUESTIONS
- 1. On Page 2 of the cover letter for the submittal, it is stated that the implementation of this amendment is not expected to require changes to the plant's Update Final Safety Analysis Reports (UFSARs). Later, the submittal states that the re-analysis of the radiological consequences of various design basis events has shown increases in the postulated consequences. 10 CFR 50.71(e) requires that the UFSAR be updated to reflect the information and analyses submitted to the Commission, including all safety analyses and evaluations performed by the licensee in support of approved license amendments. The NRC staff believes that changes will be necessary to the UFSAR. Please explain the basis for the statement that changes will not be needed for the UFSAR.
Response
The basis for the conclusion that no changes to the UFSAR will be needed is that the requested license amendment is for a lead assembly program involving only four fuel assemblies. This is consistent with past practice where Duke has placed lead assemblies in a reactor to test or verify new fuel assembly designs or design features and has not included the analyses and descriptions associated with the new fuel design in the UFSAR. Also, the radiological consequences associated with the MOX fuel lead assemblies are not significantly different from those of low enriched uranium (LEU) cores and the consequences are still well within applicable regulatory limits. Given that lead assembly programs are relatively short, fixed duration "tests" for new or revised fuel designs, there is no need to revise the UFSAR descriptions until the new fuel design is implemented.
Duke anticipates updating the UFSAR as part of the transition to batch use of MOX fuel. At that time UFSAR descriptions, including fuel assembly design, safety analyses, plant system changes, and radiological analyses, would be updated to address changes associated with operation of up to 40% MOX fuel cores. If as a result of the NRC review of this license amendment request it is determined that it would be necessary or desirable to update selected parts of the UFSAR, then Duke will make the appropriate changes in accordance with the requirements of IOCFR50.71(e).
- 2. On Page 3-2 of the submittal, the last paragraph addresses the post-irradiation evaluations to be performed on the lead test assemblies. The discussion states that these evaluations are being performed to verify the validity of the licensee's models to predict fuel assembly performance and confirm the applicability of the European database to the licensee's use of weapons-grade MOX fuel. This discussion doesn't explicitly address verification of the licensee's assumptions with respect to fission product inventory and transport within the fuel pins (gap fractions) or the impact of higher burnups on these parameters. Will the licensee be performing post-irradiation evaluations of fission product releases from the irradiated LTAs? If not, what verification wvill be performed to validate these assumptions, especially those related to (a) the extension of US experience with LEU fuels, and (b) the extension of European experience with reactor-grade MOX, to the proposed weapons-grade MOX?
86
Response
Section 3 of Reference Q2-1 addresses the issue of the applicability of the European reactor grade MOX fuel experience base to the planned use of weapons grade MOX fuel in United States nuclear power reactors. This topical report is currently under NRC review. As described in Section 8.5.3 of the Framatome ANP MOX Fuel Design Report (Reference Q2-1), current plans are to perform hot cell post-irradiation examinations of MOX fuel rods at the Oak Ridge National Laboratory. The MOX fuel will have been irradiated for three operating cycles in the Catawvba reactor, reaching burnups of 50-60 GWd/MThm. The hot cell post-irradiation examinations are expected to include measurements of fission gas release, i.e., the quantity and composition of gas in the pellet-cladding gap of the MOX fuel rods. The data from those examinations should provide additional confirmation of MOX fuel performance models that are based on the existing European MOX fuel experience base.
Reference Q2-1. BAW-10238(NP) Revision 1, MOX FuelDesign Report, Framatome ANP, May2003.
- 3. Section 3.7.3 addresses the radiological consequences of postulated design basis accidents.
The analysis descriptions provided are insufficient to support the requisite determination by the NRC staff. Please provide the following information for any radiological analysis that forms the basis for any result or conclusion stated in Sections 3.7.3, 4.2.1.3, and 5.6.3.1 in the submittal.
- a. A tabulation of analysis inputs and assumptions and their bases and justifications for Catawba or McGuire. Please provide this information in sufficient detail for the staff to perform confirmatory calculations.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response provides the requested information for Catawba only.
Analyses and Evaluations The radiological consequences conclusions were based upon a combination of evaluations and analyses. Accidents with radiological consequences were placed into two general groupings for the purpose of determining the need for evaluation or further detailed analysis; 1) accidents involving damage to a few fuel assemblies (fuel handling (FHA) and weir gate drop (WGD) accidents), and 2) those accidents which involve damage to a significant portion of the core. In the first grouping, a small number of fuel assemblies are involved such that if all of the mixed oxide (MOX) lead assemblies were in the damaged population, they would comprise all (or a significant portion) of the population. In the second grouping, the amount of predicted core damage ranges from 11% (locked rotor) to half core (rod ejection accident) to full core (LOCA).
Since these accidents involve a relatively large amount of fuel, the relative effect of damaging all four MOX fuel lead assemblies is reduced as the fuel damage population increases. In the cases where there is broader core damage, the effects of the four MOX fuel lead assemblies can be evaluated based upon the results of the radiological analyses for full low enrichment uranium (LEU) cores and the projected increase in I-131 inventory and release for a MOX fuel assembly relative to an LEU fuel assembly. In a large population of failures, the impact of the MOX lead fuel assemblies is small.
87
Accident Evaluations With four MOX fuel lead assemblies included in the percentage of affected fuel, the contribution of these assemblies to the overall impact of the accident decreases as the number of failed fuel assemblies increases (Catawba cores contain 193 fuel assemblies). Where failures occur in a limited number of fuel assemblies, the overall impact of the release will be governed by the fuel type of the majority of the assemblies affected, i.e., LEU fuel assemblies. The radiological impact of the accident can be evaluated by examining the limiting dose for the scenario. The limiting dose is typically the thyroid dose which is driven by the I-131 release. The increase in I-131 released from a MOX fuel assembly relative to an LEU fuel assembly can be computed.
By dividing the four MOX fuel lead assemblies by the total population of affected fuel pins and applying the percentage of I-131 release increase for MOX fuel, the thyroid dose result for an all LEU core can be used to calculate a predicted dose with the four MOX fuel lead assemblies included in the affected fuel population. This approach conservatively assumes that all fuel pins in all four MOX fuel assemblies experience fuel failure, along with some (or, in the case of LOCA, all) of the other (LEU) pins in the core. However, because the MOX fuel lead assemblies are required by the Catawba Technical Specifications to be in non-limiting locations, the MOX fuel assemblies may not experience fuel failure. With no MOX fuel pin failures, no impact on dose consequences would be observed.
Ratio of 1-131 for MOX and LEU Fuel for Accident Evaluations Radiological inventory calculations for MOX fuel and LEU fuel indicate that the concentrations of radionuclides that impact dose are generally similar. The differences vary from isotope to isotope. Because thyroid doses are typically the limiting dose in radiological accident analysis, a comparison of I-131 inventories is considered to be an appropriate means of estimating relative impacts from MOX and LEU fuel.
Isotopic inventories vary as a function of burnup and plutonium concentration (for MOX fuel) and uranium enrichment (for LEU fuel). Section 3.7.3 of Attachment 3 to Reference Q3(a)-1 cites an increase of about 3% in the I-131 inventory of MOX fuel relative to LEU fuel. This is based on a plutonium concentration of 4.37% (nominal assembly overall average) and a uranium enrichment of 4.0% (approximately equivalent from the standpoint of reactivity over two cycles).
This comparison was made at a burnup of 45 GWd/MThm, which would be characteristic of assemblies that are operated at relatively high power for two cycles.
However, it is recognized that the MOX fuel assembly is radially zoned so it will actually have three different plutonium concentrations (see Section 5.3 of Reference Q3(a)-2). For an assembly with a nominal average plutonium concentration of 4.37%, the actual concentrations are 2.40% (12 pins), 3.35% (76 pins), and 4.94% (176 pins). For MOX fuel, I-131 inventory increases with initial plutonium concentration. In addition, while I-131 concentration for LEU fuel increases continually with burnup, the I-131 concentration for MOX fuel peaks early in life and then decreases continually. In order to make a more conservative estimate of relative impact of MOX fuel, the I-131 concentration for 5% plutonium MOX fuel at 16.9 GWd/MThm was compared to the I-131 concentration for 4% U-235 fuel at 60 GWd/MThm. 16.9 GWd/MThrn is considered to be a bounding low burnup for the MOX fuel lead assemblies at the end of the first cycle of irradiation, and 60 GWd/MThrn is the maximum allowable burnup for LEU fuel at 88
Catawba. These are considered to be the worst cases for a fuel handling accident for representative assemblies over the bumup ranges that might be experienced by a three-cycle MOX fuel lead assembly. Using this approach, the inventory of I-131 in the MOX assembly is calculated to be about 9% higher than that in the equivalent LEU assembly. For loss of coolant (LOCA) and departure from nucleate boiling (DNB) accidents, four fuel assemblies are a small portion of the projected fuel failures in these scenarios, so the impact of the failure of all four MOX fuel lead assemblies in these scenarios is also projected to be small, as shown in the response to radiological question 3(b).
Accident Analyses The methodology utilizing the MOX/LEU I-131 ratio could also be applied to the fuel handling accidents. However, because these accidents deal with small numbers of failed assemblies such that the four MOX fuel lead assemblies could theoretically populate all or the majority of the affected assemblies, it was considered prudent to explicitly analyze these scenarios. The fuel handling accidents that are currently analyzed at Catawba are the single assembly drop fuel handling accident and the weir gate drop (WGD) accident. The WGD involves seven fuel assemblies. For the WGD analysis, all four MOX fuel lead fuel assemblies were assumed to be in the population of damaged fuel assemblies. The other three damaged fuel assemblies are LEU. The single assembly drop and WGD analyses were performed in accordance with Alternative Source Term (AST, Reference Q3(a)-3) technology, which reflects the current Catawba licensing basis for fuel handling accidents.
Additionally, a fuel handling accident scenario involving damage to a fresh MOX fuel assembly was analyzed for off-site and control room impact. This methodology and model are similar to those used in the evaluation of this scenario at the MOX Fabrication Facility (References Q3(a)-
4 and Q3(a)-5). See the response to radiological consequences question 3(e) and the discussion of inputs to this analysis below for further information.
MOX Spent Fuel Handling Accident and Weir Gate Drop Accident Model Inputs For the analyses involving spent fuel, except for the fuel assembly isotopics, these models are basically the same as the FHA and WGD models that were reviewed for the Catawba FHA AST submittal in Reference Q3(a)-6. This submittal was approved in a subsequent NRC Safety Evaluation in Reference Q3(a)-7. The differences between that submittal and the analyses performed in support of the MOX lead fuel assembly submittal are discussed in the Table Q3(a)-
1 below; all other features of the model are the same as Reference Q3(a)-6.
MOX Fresh Fuel Assembly Damage Model Inputs In this scenario, a fresh MOX fuel assembly is dropped and damaged during handling prior to its placement in the spent fuel pool. An off-site exclusion area boundary dose and a control room operator dose were calculated. Dose criteria as discussed in the response to Reactor Systems Question 12 were adopted. This analysis was performed to be applicable to both McGuire Nuclear Station and Catawba Nuclear Station since it was performed prior to the decision to remove McGuire from consideration to irradiate the lead assemblies. Therefore, the analysis was performed using values from both stations. These values were chosen to bound the parameters at either plant.
89
This scenario is considered to be a variation of the spent fuel FHA. The methodology and modeling used for this analysis were consistent with the methodology and modeling used in Duke Cogema Stone & Webster (DCS) calculations supporting the MOX Fuel Fabrication Facility Construction Authorization Request (Reference Q3(a)-4). NRC Office of Nuclear Material Safety and Safeguards reviewed the DCS calculations and found them to be acceptable (Reference Q3(a)-5). Significant inputs for this analysis are summarized in Table Q3(a)-2.
See responses to Reactor Systems Question 12 and Radiological Consequences Questions 3(e),
3(f) and 3(g) for further information. See response to Radiological Consequences Question 3(b) for results.
References Q3(a)-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.
Q3(a)-2 BAW-10238(NP) Revision 1, MOX Fuel Design Report, Framatome ANP, May 2003.
Q3(a)-3 USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q3(a)4 Letter from DCS to USNRC, DCS-NRC-0001 14, "Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization Request Revision," CAR Revision 1, October 31, 2002.
Q3(a)-5 Letter USNRC to DCS, "Draft Safety Evaluation Report On Construction Of Proposed Mixed Oxide Fuel Fabrication Facility," Revision 1, April 30, 2003.
Q3(a)-6 Letter from G. R. Peterson (Duke) to USNRC, "Proposed Amendment for Partial Implementation of Alternate Source Term and Proposed Amendment to Technical Specifications (TS) 3.7.10, Control Room Area Ventilation System, TS 3.7.11, Control Room Area Chilled Water System, TS 3.7.13, Fuel Handling Ventilation Exhaust System, and TS 3.9.3, Containment Penetrations," December 20,2001.
Q3(a)-7 Letter from C. P. Patel (USNRC) to G. R. Peterson (Duke). "Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments," April 23,2002.
90
Table Q3(a)-1 Significant Inputs for the Spent Fuel Weir Gate Drop and Fuel Handling Accident Models Parameter Discussion;1' The source term involves the usage of MOX fuel. Detailed discussion of this Source Term source term and its development are contained in the responses to reactor systems question 34 and radiological consequences question 3(f).
Standard AST release fractions from Reference Q3(a)-1 and Reference Q3(a)-
3 were used. Additionally, these release fractions were raised to investigate Release Fractions their sensitivity in view of the increased fission gas release in the MOX assembly. Detailed discussion of the release fractions used is contained in response to radiological consequences question 3(g).
Decontamination The more restrictive requirement of Reference Q3(a)-3 is to model an overall Factors DF of 200 for the spent fuel pool or fuel transfer canal water. With an organic DF of 1, an elemental DF of 350 was calculated.
Control room ventilation is assumed to be unbalanced with 60% of the flow into the control room being drawn from a contaminated stream. Therefore, the Control Room X10 control room X/Q value from Reference Q3(a)-1 is inci eased to a value of 1.04E-3 seclm3 for Unit Vent releases.
91
Table Q3(a)-2 Significant Inputs for Fresh Fuel Damage Models Parameter,
."Disc--ssin
-'2:.
The damage ratio used was 0.01. This means that 1 % of all of the fuel pellets Damage Ratio in all of the fuel rods in the assembly are damaged from the fall. This value is applicable to falls from heights up to 30 feet.
Based upon the elevations in the spent fuel pool area related to the lifting of a Height of Drop fuel assembly, the maximum height that the assembly could drop was set to 30 feet.
Airborne All isotopes released, as a result of the drop, which become airborne will be Respirable Fraction inhaled. A conservative value of 1.0 is used for this fraction.
This value models the portion of the fuel that becomes airborne from the Release Fraction fraction of the fuel that is damaged. A value of 1.96E-4 was calculated based upon curve fits to experimental data.
Plutonium concentration:
5%
MOX assembly loading: 0.4626 MThm Pu 238 % of MOX:
0.025%
Pu 239 % of MOX:
92.50%
Pu 240 % of MOX:
6.925%
Source Term Pu 241 % of MOX:
0.50%
Pu 242 % of MOX:
0.05%
U 234 % remainder of fuel:
0.0017%
U 235 % remainder of fuel:
0.25%
U 236 % remainder of fuel:
0.0012%
U 238 % remainder of fuel:
99.7471%
Credit is taken for only one set of filters in the flow path from the spent fuel Filtration pool to the atmosphere and the control room. A conservatively low particulate efficiency of 95% was modeled.
A bounding value of 9.OE-4 sec/iM3 was used. Note that this is a McGuire Unit Vent X/Q value. It bounds the Catawba value of 5.5E-4 sec/M 3.
Control Room A bounding value of 1.74E-3 sec/M 3 was used.
x/Q 92
Question 3:
- b. Provide the numeric results of the analyses discussed in Sections 3.7.3, 4.2.1.3, and 5.6.3.1 in terms of the whole body and thyroid dose quantities, or total effective dose equivalent (TEDE), as appropriate to the licensing basis of Catawba and McGuire.
Include offsite and control room doses.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response provides the requested information for Catawba only.
Evaluations As was discussed in the response to Radiological Consequences Question 3(a), the thyroid dose is typically the limiting dose in that there is less margin between calculated thryroid dose and thyroid dose limit than between calculated whole body dose and whole body dose limit. In reviewing the margins to limit for the existing TID (Reference Q3(b)-
- 1) analysis, the loss of coolant (LOCA) scenario has the least margin to the limit for both whole body and thyroid. For LOCA whole body doses, 80% of the margin to the limit exists, and for thyroid doses 70% of the margin to the limit currently exists. Since the LOCA is the most restrictive accident and its thyroid dose is more restrictive than its whole body dose (although margin to the limit exists for both), the evaluation of the TID based accidents is performed based upon the thyroid doses for these accidents.
As also discussed in the response to Radiological Consequences Question 3(a), all of the mixed oxide (MOX) fuel lead assemblies are assumed to be in the affected fuel population for all of the accidents examined. Thus, for those scenarios where a significant portion of the core is affected, the increase in dose is calculated by taking the relative change in 1-13 1 multiplied by the number of affected MOX lead assemblies (4),
divided by the total population of the 193 fuel assemblies in the core affected. A 50%
increase in the TID or AST (Reference Q3(b)-2) release fractions (discussed in the response to Radiological Consequences Question 3(g)) for the MOX fuel lead assemblies is modeled by multiplying by 1.5. For the LOCA, all fuel assemblies are assumed to be affected; for the locked rotor accident, 11% of the core is assumed to be affected; and for the rod ejection accident 50% of the core is assumed to be affected.
Because the existing analysis is based on a full core of low enrichment uranium (LEU) fuel assemblies, the evaluation is concerned with quantifying the increase in the dose result from replacing four LEU fuel assemblies with four MOX fuel lead assemblies.
The dose result already includes the effects of the four LEU fuel assemblies, so the evaluation quantifies the increase from the replacement by accounting for the difference in the effects of these two types of fuel assemblies. It is this difference that results in the change in dose.
The LOCA evaluation is illustrated as follows:
- The four MOX fuel lead assemblies are divided by the number of (193-- full core failure) fuel assemblies failed, to yield the percentage of the failed population 93
comprised of the four MOX lead assemblies.
(Fraction of the population of failures that are MOX assemblies = 4/193)
The MOXILEU I-131 ratio provides the increase in I-131 due to the replacement of an LEU fuel assembly with a MOX fuel assembly.
(As discussed in the response to Radiological Consequences Question 3(a), the increase of I-131 for MOX relative to LEU is 9%)
To conservatively account for the increased gas fission product release to the fuel-cladding gap a bounding safety case for the MOX fuel lead assemblies is desired.
The increased fission product release to the fuel-cladding gap is modeled by increasing the release fractions by 50% (see response to Radiological Consequences Question 3(g)). This is accomplished by multiplying by a factor of 1.5.
The values in the three bullets above are multiplied to get the percentage increase in the thyroid dose from the MOX lead assemblies ((4/193)
- 9%
- 1.5 = 0.28%).
The percentage increase of the thyroid dose is multiplied by the existing analysis result (as shown in Table Q3(b)-1) to calculate the increase in the dose due to the MOX lead fuel assemblies. It is then added to the existing result to calculate the projected thyroid dose with the MOX lead assemblies. For the EAB thyroid dose for the TID LOCA scenario, the current dose is 89 Rem.
(0.28%
The same calculational process yields an increase in the locked rotor thyroid dose of 2.5% and the rod ejection thyroid dose of 0.6%. The respective total projected exclusion (EAB) thyroid doses are 3.8 Rem and 1.01 Rem. All of these increases are very small.
These results are summarized in Table Q3(b)-2. The results do not remove an appreciable amount of margin.
Analyses The impacts to doses in the evaluations described above are small due to the low number of MOX fuel lead assemblies as a portion of the population affected. The fuel handling accident scenarios provide for the most impact because MOX fuel comprises most or all of the fuel that is assumed to be damaged. The single assembly fuel handling accident (FHA) and the weir gate drop (WGD) accident were analyzed. For the purpose of comparison, the doses were computed using a MOX fuel source term and an equivalent LEU source term (see response to Radiological Consequences Question 3(f) for further discussion of equivalent MOX and LEU fuel assemblies). The unit vent is the limiting release point.
As discussed in the response to Radiological Consequences Question 3(g), increases in AST release fractions were examined to determine the sensitivity to these values. These and other results of the analyses for MOX and LEU for the FHA and WGD are presented in Tables Q3(b)-3 and Q3(b)-4. Included in these tables is an evaluation of the results for the Catawba AST LOCA including MOX lead assemblies. This analysis is currently under NRC Staff review (Reference Q3(b)-3).
94
For the FHA, consequences increased by about 9%. This matches the single assembly FHA increase in I-131 activity for MOX fuel relative to LEU (see response to Radiological Consequences Question 3(a) "Ratio of I-131 for MOX and LEU Fuel for Accident Evaluations"). Because the WGD involves both MOX and LEU fuel, the effect of the MOX fuel is reduced. The increase in I-131 for the WGD accident was 5% and the increase in dose results was approximately the same. Additionally, when the MOX release fractions were increased to account for differences in gap fractions, the impact on the results (relative to MOX fuel analyses with standard release fractions) was an increase in similar proportion to the 1-131 release fraction. These results demonstrate that methodology involving the use of an I-131 activity ratio for evaluating accident consequences (as discussed and demonstrated above) provides valid results.
The dose consequences of the fresh fuel assembly drop analysis are reported in Section 3.7.3.5 of Attachment 3 to Reference Q3(b)-4.
See responses to Reactor Systems Question 12 and Radiological Consequences Questions 3(a), 3(f) and 3(g) for additional information.
References Q3(b)-1 USAEC, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.
Q3(b)-2 USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q3(b)-3 Letter from G. R. Peterson (DPC) to USNRC, "Proposed Technical Specifications and Bases Amendment: Technical Specification and Bases 3.6.10 Annulus Ventilation System, Technical Specification and Bases 3.6.16 Reactor Building, Technical Specification and Bases 3.7.10 Control Room Area Ventilation System (CRAVS), Technical Specification and Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES),
Technical Specification and Bases 3.7.13 Fuel Handling Ventilation Exhaust System (FHVES), Technical Specification and Bases 3.9.3 Containment Penetrations, Technical Specification and Bases 5.5.1 Ventilation Filter Testing Program (VFTP)," November 25, 2002.
Q3(b)-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.
95
Table Q3(b)-1 Offsite and Control Room Doses with LEU Cores and Projected Doses with MOX Lead Assembly Cores for LOCAs with TID and AST Releases 1 Catawba UFSAR 2 Reference Q3(b)-3 Table Q3(b)-2 Offsite Thyroid Doses with Full LEU Cores and Projected Thyroid Doses with MOX Lead Assemblies for Locked Rotor and Rod Ejection Accidents 1 Standard TID releases.
2 Increased TID releases.
3 Note that for those accidents where there are multiple applicable cases (such as for concurrent and pre-exiting spiking), the result of the case with the highest dose is shown.
96
Table Q3(b)-3 Weir Gate Drop Accident TEDE Dose Results with Unit Vent Release Over 2 Hours MOX Fuel with Percentage Receptor Dose Pecnae Increased Increase Limit D
LEU Fuel MOX Fuel Increase Release Increase Limit MOXILEU Fractions Increased MOX (Rem)
(Rem)
(Rem)
(Rem)
Releases EAB 6.3 2.2 2.3 5%
3.5 58%
LPZ 6.3 0.31 0.33 5%
0.50 58%
Control 5.0 2.1 2.2 5%
3.3 58%
R o o m Table Q3(b)-4 Fuel Handling Accident TEDE Dose Results with Unit Vent Release Over 2 Hours 97
Question 3:
- c. If the analyses utilized atmospheric dispersion coefficients (X/Q) that were generated in support of this amendment request, please describe the assessment method used and provide all inputs and assumptions used in the assessment. Please provide a computer readable copy (e.g., CD) of site meteorological data used in the assessment.
Response
No new atmospheric dispersion factors were generated in support of this amendment request. See response to Radiological Consequences Question 3(a) for more discussion of the specific values used.
Question 3:
- d. Provide the requested information item 1(a) of Generic Letter 2003-01, "Control Room Habitability." Please note that the response to this particular question does not relieve the licensee of the responsibility of responding to Generic Letter 2003-01.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response provides the requested information for Catawba only.
Catawba Nuclear Station control room inleakage testing has been conducted. The response to Generic Letter 2003-01 will be provided by December 9, 2003.
Question 3:
- e. If the analyses used methods, inputs, or assumptions different from that in the current licensing basis (CLB) for Catawba or McGuire, provide ajustification for each change from the CLB.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response addresses Catawba only.
As discussed the response to Radiological Consequences Question 3G), the current source term licensing basis for Catawba is transitioning from classical source term modeling (TID, Reference Q(e)-1) to Alternative Source Term (AST, Reference Q(e)-2), but the evaluations and analyses that were performed in support of the mixed oxide (MOX) fuel lead assemblies submittal were in keeping with the current licensing basis (CLB) of Catawba at the time it was submitted. The only difference between the current licensing basis analyses and the MOX fuel analysis for the spent fuel handling accidents is the use of MOX fuel, which is the subject of the license amendment request (LAR). Methods, inputs and assumptions not directly related to MOX fuel were not changed for the evaluations and analyses to accommodate MOX fuel lead assemblies (see response to Radiological Consequences Question 3(a)). The models used in the evaluations and 98
analyses of accidents with MOX fuel lead assemblies were based upon those in place for the analyses supporting the Catawba CLB.
Formal analyses were performed for the single spent fuel assembly accident and the weir gate drop accident (involving seven fuel assemblies). Evaluations were performed for those accidents where the four MOX lead assemblies would account for less than 20% of the damaged fuel population (see responses to Radiological Consequences Questions 3(a) and 3(b) for further information). These evaluations were based upon the comparison of isotopics between a MOX and an equivalent low enrichment uranium (LEU) fuel assembly (see further discussion in response to Radiological Consequences Question 3(f)).
Catawba is currently licensed to use Alternative Source Term (Reference Q3(e)-2) technology for fuel handling accidents. AST-based MOX fuel release calculations were performed to provide a baseline analysis using MOX fuel source terms and AST gap fractions. Cases were also run with a "raised" AST-based source term. This source term increased the release fractions but maintained the AST isotopic release distribution. The raised AST release analyses were performed to provide a bounding safety case of higher gap release fractions for the MOX fuel lead assemblies. The use of these gap fractions cases are discussed in response to Radiological Consequences Question 3(g).
There is no formal analysis of a dropped fresh LEU fuel assembly in air for Catawba.
For MOX fuel it is appropriate to analyze this variation of the FHA because the potential for dose consequences are more severe, due to the presence of plutonium in the fuel pellets.
See responses to Radiological Consequences Questions 3(a), 3(b), 3(f), 3(g), and 3(i) for further information.
References Q3(e)-1 USAEC, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.
Q3(e)-2 USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Question 3:
- f. Please provide a tabulation of the isotopic core inventory, in Curies (Ci) or micro-Curies (uCi). Please identify the computer code used to determine these values.
Please explain the derivation of the fuel-specific code libraries used in this assessment. If LEU libraries were used, please provide a scrutable justification of why the results should be considered applicable to the proposed MOX LTAs.
Response
The development of the mixed oxide (MOX) source isotopics was discussed in the response to Reactor Systems Question 34. An industry standard code package, SCALE, was used to generate the MOX source term isotopics. These isotopics were used as the source term in the accident analyses and evaluations that were performed, including 99
evaluations of the potential impact from a MOX fuel assembly relative to a similar low enrichment uranium (LEU) fuel assembly (whose isotopics were determined using the same computer codes and processes as the MOX assembly).
The general accident scenarios for which source terms were developed were the fuel handling accidents, the loss of coolant accidents, and departure from nucleate boiling (DNB) related accidents. SCALE/SAS2H cases were constructed to model combinations of these accident scenarios, MOX fuel plutonium concentrations, amounts of fuel assembly bumup, and burnup schemes. The models were built to include conservatisms.
The isotopic data were compiled and tabulated for the development of source terms which would be specific to the scenarios and the needs of future analysis requirements.
Fuel Assembly Modeling Variables MOX fuel plutonium concentrations and assembly bumup ranges were specified based upon projected MOX fuel usage and the plant licensing basis. The range of plutonium concentrations was further divided into even increments so as to provide data for future evaluations and usage. The maximum burnup range was expanded to include several possible maximum bumup points. These "end of life" burnups were used to establish a cyclic burnup scheme for each requiring the calculation of isotopic inventories at the end of each burn cycle. This level of detail resulted in many statepoints of plutonium concentration and assembly burnup. The combinations of all of these statepoints, plus those from other variables resulted in numerous cases.
Peaking factors were developed for three different scenarios since the peaking is different in each: loss of coolant accidents (LOCA), departure from nucleate boiling accidents, and fuel handling accidents (FHA, including the weir gate drop accident). Since the FHA scenarios involve only a limited number of fuel assemblies, a bounding peaking factor, rather than lead assembly core-specific peaking factors, was specified.
The LOCA and DNB peaking factors were computed based upon projected fuel duty.
For the LOCA scenarios, core design predictions for three MOX fuel lead assembly cycles were utilized for burnup parameters. The projected lead assembly bumup produced peaking factors for each cycle. These peaking factors were used to compute cycle specific power history parameters for-the SCALE models. For the DNB scenarios, a bounding lead assembly power history (from beginning to end of life discharge bumup) for DNB accidents was taken from the core design predictions. These peaking factors were used to compute the power history parameters in the SCALE models for each cycle.
In the FHA and WGD scenarios, peak inventory assemblies are involved in the accident.
A peaking factor of 1.65 was used in the SCALE models. This value was taken from Reference Q3(f)-1 and is a standard assumption in FHA analyses. This value is above the peaking factors calculated in the LOCA and DNB cases. It is also greater than the typical Catawba core maximum radial peaking factor of 1.40 to 1.50, satisfying the requirements of Reference Q3(f)-2. Therefore, a peaking factor of 1.65 is used in the MOX FHA cases in computing the cycle specific power history parameters in the SCALE models.
100
Equivalent MOX and LEU Fuel Assemblies In order to make comparisons between the impacts of a MOX fuel assembly (the Mark -
BW/MOX1 design) and an equivalent LEU fuel assembly (the Westinghouse Robust Fuel Assembly design), it is first necessary to define the parameters that will equate these two assemblies. These parameters are assembly burnup, LEU enrichment, and MOX fuel plutonium concentration. The MOX fuel assembly contains 264 fuel pins. The pin population of the assembly is comprised of three groups of pins. Each group has a different plutonium concentration. This results in an assembly averaged plutonium concentration of 4.37% for the current nominal design. Source term analyses are typically performed in terms of assembly averaging since the codes involved perform the calculations on an assembly, rather than individual fuel pin, basis.
In evaluating the isotopics for spent fuel, the intended use of the source term must also be a consideration. Since analyses were conducted for the fuel handling accidents due to the potential impact of the MOX fuel assembly on the small population of fuel assemblies involved in the accident, this scenario was established as the basis for isotopic evaluation.
This is also the most straight forward comparison since the peaking factors are the same and constant for both the LEU and MOX fuel source term models. Historically, the limiting dose in these scenarios has been the dose to the thyroid. The thyroid dose is mainly driven by the release ofI-131. It can be reasoned that the doses will continue to be dominated by the impact to the thyroid and, in particular, by the release of I-131.
Therefore, the source term results can be compared for dose impact by comparing the magnitude of 1-131 available for release. A review of the source term cases yielded several conclusions related to MOX spent fuel isotopics:
- I-131 increases with increased plutonium c3ncentration (although it shows small variations over the concentration range of interest). Therefore, a higher plutonium concentration will yield a slightly higher amount of 1-131 and slightly higher doses.
The I-131 inventory in MOX fuel peaks early in life and then decreases with bumup.
From the above it is concluded that the MOX source term will use an assembly averaged plutonium concentration of 5.0% (based upon a nominal maximum plutonium pin concentration of 4.94%) MOX with a burnup of 16.9 GWd/MThm. This burnup bounds the lowest expected burnup during the first refueling outage, so this source term will maximize the I-13 1 inventory for a MOX fuel lead assembly fuel handling accident.
For the purposes of comparison it is desired to define an "equivalent" LEU fuel assembly. The two variables of interest are enrichment and burnup, similar to the MOX assembly. As was done for the MOX fuel assembly, the bumup for the LEU fuel assembly is that bumup which produces the largest magnitude of I-131 inventory.
Review of the LEU isotopic cases shows that I-131 increases with increasing bumup.
The maximum bumup that Catawba is licensed to currently is 60 GWd/MThm.
The equivalent enrichment is based upon reactivity. There is some dependence on the range of burnup, but typically an LEU enrichment of 4.10-4.20% has been considered to 101
be equivalent to a MOX fuel plutonium concentration of 4.37%. While the MOX fuel assembly modeled is based upon a fuel assembly with an average plutonium concentration of 4.37%, the maximum pin concentration was adopted for conservatism.
Review of LEU isotopic data shows while 1-131 inventory increases with decreasing initial uranium enrichment, the change in I-13 1 magnitude over the range of enrichments is small. Therefore, 4.0% is considered to be an appropriate enrichment for the equivalent LEU assembly.
Two tables of isotopic composition follow. Table Q3(f)-I shows the isotopic inventory for a 5.0% MOX fuel assembly subject to fuel handling accident peaking factors and a bumup of 16.9 GWd/MThm. Table Q3(f)-2 shows the isotopic inventory for an "equivalent" 4.0% enriched LEU fuel assembly with a bumup of 60 GWd/MTU.
See the responses to Radiological Consequences Question 3(a) and Reactor Systems Question 34 for additional information.
References Q3(f)-1 Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," Revision 0, March 1972.
Q3(f)-2 USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
102
Table Q3(f)-1 Single MOX Fuel Assembly Isotopic Inventory:
5.0% MOX Concentration, 16.9 GWd/MThm Burnup and Fuel Handling Accident Peaking iBarium, Strontium Noble Gases Halogens Alkali Mfetals Tellerlum Group Group Noble M etals Lanthanides Cerium Group Actinide Group (C I)
)
(C 1 (CI)
)
(C 1(
))
(C
(
)
(C I)(CI kr 3m 7.25E+04 br S3
- .22E+04 rb t6 3.13E+02 Be tl 3.77E+04 or tg 4.31 E1+005 mo 99 1,4tE+06 y 90 9.72E+03 ce141 1.25E+06 u235 I. 1E.03 kr S5m 1.34E+05 br15 1.33E+05 rb 3t 3.4tE+05 se tl 3.79E+04 sr 90 9.59E+03 molOI 1.42E+06 Y 91 6.15E+05 ce143 1.0oe+06 u237 1.13E+05 kr 35 1.39E+03 br 37 1.38E405 rb t9 4.39E+05 rt t3m 3.29E+04 ir 91 6.37E205 molO2 1.43E+06 y 91m 3.71E+05 ce144 4.42E+05 u238 1.46E.01 krt 7 2.52E+05 1130 7.51E203 rb 90 3.72E+05 se 4 1.13E+05 sr92 7.49E+05 te 99m 1.30E+06 y 92 7.53E+05 ce145 7.45E+05 u239 1.35E+07 kr t 3.35E+05 1131 8.tlE+05 cs234 1.94E+04 sei 7 5.61E+04 sr93 9.21E+05 tclOI 1.42E+06 y 93 6.31E+05 ce146 6.12E+05 kr t9 3.t4E+05 i132 1.2tE+06 cs136 3.41E+04
- bl27 1.02E+05 ba139 1.36E+06 tc104 1.40E+06 y 94 1.06E+06 np237 2.09E202 xel31m 1.07E+04 1133 1.64E+06 crl37 2.67E+04
- b12S 1.77E+04 bal40 1.36E+06 ru103 1.60E+06 Y 95 1.15E+06 np233 2.57E+04 xel33m 5.51E+04 1134 1.76E+06 cs138 1.43E+06 sbl21m 1.50E+05 ba241 1.24E+06 ru05 1.26E+06 zr95 1.IIE+06
,,p239 1.34E+07 xc133 1.65E+06 1235 1.57E+06 cs139 1.31E+06 sb129 3.21E+05 ba242 1.13E+06 rutl06 3.ttE+05 zr97 1.24E+06 np 2 40 4.47E+04 xel35m 3.92E+05 sb 130 1.16E+05 rul07 7.89E+05 nb 95 1.04E+06 pu236 1.32E202 xet35 6.t0e+05 abl30m 3.53E+05 rblO3m 1.59E+06 nb 95m 1.23E+04 pu23S 1.66E+02 l
xe137 1.48E+06
- bl31 6.1iE+05 rhbOS5 1.123E+06 ib 97 1.25E+06 pu239 9.31E+02 xe133 1.26E+06 b132m 3.1tE+05 rblO6m 2.31E+04 bl40 1.37E+06 pu240 S.0OE+02 1e127m 1.47E+04 rhl07 7.91E+05 2a14t 1.26E+06
_ pu241 1.49E+05
_te27 9.36E+04 pdIO9 4.60E+05 Ws142 1.20E+06 pu242 5.1tE-01
.e229 3.06E+05 pdl 12 7.25E+04 1sa43 1.07E+06 pu243 1.55E+05 1te229m 6.32E+04 pd12 3.21E+04 nd147 5.02E+05 te131 7.04E+05 ndl49 3.13E+05 2e132 1.22E+06 ndISI
- 1. 5E+05
_1e33 3.23E+05 pm147 6.15E+04 tel33m 6.94E+05 pml4t 3.53E+04 e_
1l34 1.21E+06 ppmi 4m 1.39E+04
~~P
.149 4.202+05 s ili 553 2.45E+ 05 m lS6 2.7 32+04 eu154 1.32E+03 eU255 3.28E+02
= =_
_ =_ = = =
=
=
s_ 56 7.90E+04
=
=
=
ctot ?
2.202+04
~~~~~~
~ ~~~~pr242 1.122 +04 pr2 43 2 0 7 2 + 0 6
_ _ _ _ _ __1
_07 E_0 6
~~~~~~
~~~~ ~~ ~~pi144 4.492+05
~~ ~~pr244m 6.26 2+03
~~~prI4 5
7.45 2+05 pr246 6.2 0 + 0 5
_____p r_4 6_6_2 0__0_
~~~pr147 5.0 72+ 40 5 am241 7.63E+02 m 242m 4.462+ 00_
M 242 3.45 2+04
=__________
= - =_______
=_________
=_________
am243 3.72e+00
- o *
=
==-
am 244 2.73 2+04 m 242 3.7 32+03 cm244
,2.34E+02 103
Table Q3(f)-2 Single LEU Fuel Assembly Isotopic Inventory:
4.0% U235 Enrichment, 60 GWd/MTU Burnup and Fuel Handling Accident Peaking Barium, Stronlium Noble Cases Halogens AlkalI M etals Tellerlum Group Group Noble M etals Lanthanldes Cerium Group Actinide Group (C l)
(I(C C
)
(C l)
(C I)
_(
C 2 (C ' )
(C n krt_3m 7.8_1E+04 br83 7.69e+04 rbt66 2.54E+03 se 3.52E+04 sr9 5.51E+05 mo99 1.47E+06 y90 6.34E+04 cee41 1.25E+06 u235 3.79E-03 kr S5m 1.4tE+05 br85 1.47E+05 rb 81 4.08E+05 set3 3.77E+04 sr90 5.S7E+04 molOI 1.38E+06 Y 91 7.63E+05 cee43 1.12E+06 u237 9.53E+05 kr t5 6.9tE+03 brS7 2.27E+05 rb 39 5.20E+05 se 3m 3.75E+04 sr 91 7.26E+05 mlOM2 1.36E+06 y 91m 4.21E+05 ce144 9.66E+05 u23t 1.40E-01 kr 87 2.92E+05 il30 3.67E+04 rb 90 4.65E+05 set4 1.34E+05 sr92 t.16E+05 tc 99m 1.31E+06 y92 8.21E+05 ce145 7.72E+05 u239 1.95E+07 kr tS 3.96E+05 i131 t.06E+05 cs134 2.01E+05 se t7 7.36E+04 sr93 9.70E+05 telO 1.38E+06 y 93 6.60E+05 ce146 6.30E+05 kr 39 4.72E+05 1132 1.17E+06 cs136 5.833+04 sb127 3.19E+04 ba139 1.37E+06 t1C04 1.26E+06 y 94 1.03E+06 np23 7 2.99E.01 xet31m 1.23E+04 i133 1.61E+06 eu137 9.22E+04 0b128 1.37E+04 baI40 1.3$E+06 ru103 1.43E+06 y 95 1.16E+06 np23S 6.63E+05 xel33m 4.962+04 i134 1.74E+06 cs138 1.44E+06 sbl2tm 1.2tE+05 baI41 1.23E+06 rul05 1.12E+06 zr95 1.21E+06 np23 9 1.94E+07 xe133 1.54E+06 i135 1.55E+06 cs139 1.33E+06 rb129 2.76E+05 ba142 1.15E+06 ru106 5.8tE+05 zr97 1.22E+06 ap240 9.33E+04 oct35m 3.59E+05 sbl30 9.22E+04 rul07 7.08E+05 nb 95 1.22E+06 pu23 6
S.23E.0I xe135 2.35E+05 sbl30m 3.4SE+05 rhl03m 1.43E+06 nb 95m 1.34E+04 pu23S 3.78e+03 l 137 1.47E+06 sblil 6.0t1+05 hIO5 9.55E+05 nb 97 1.24E+06 pu239 2.09e+02 xe13t 1.31E+06 sbl32m 3.44E+05 rhlO6m 5.17E+04 1a140 1.53E+06 pu240 3.31E+02 tel______
3j127m 1.34E+04 rh07 7.10E+05 WU41 1.25E+06 pu2 4 1 9.98e+04 te127 3.30E+04 pdlO9 4.79E+05 1W142 1.20E+06 pu242 2.35E+00 tel29 2.63E+05 pdl 11 6.34E+04 13143 1.l IE+06 pu243 7.90e+05 tel29m 5.33E+04 pdll2 2.93E+04 ndl47 5.16E+05 tel____
31 6.71E+05 ndl49 3.26E+05 te132 1.15E+06
_ndl51
__IdE+05 te333 t.48E+05 pm 147 9.15E+04 tel33m 6.922+05 pm 148 1.54E+05 te134 1.31E+06 pml48m 1.57E+04 pm149 5.t7E+05 pm 151 1.t2E+05 am 153 6.06E+05 l I
_m 156 2.52E+04
.__=______
eu154 8.20e+03
==
=
=
__________CU I55 3.14E+03 eul56 4.11 2e+05 E_05
_eu57 5.49E+04 prl42 9.522+04
_______9 2
p~~~~~
~~~~~~~~~ ~ ~~ ~~r143 I 1.08 2+06 p.144 9.302+05 9.8 0 E + 0 prl44m 1.36E+04 prI 4S 7.732+ 5
________r14 5
7.7 3 E_0 5 pr3 4 6 6-.376 2E++05 __________
prl47 5.132+05 E
_0_
sm241 7.23e+01 a~~~~~~ ~~~~m242m 4.692+00 amm22422m 7.465 2+04 __________
7 E+04 a~~~~~
~~~~~~~m243 3.942+01 a~~~~~ ~ ~~~~~~~m244 4.152E+05
=_________ __________
=_____
mcm 242 4.19E+04
=_________=======___
====______
cm244 9.32E+03
==
104
Question 3:
- g. The CLB fuel handling and weir gate accidents at Catawba are based on the alternative source term and were performed in accordance with the guidance of Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Footnote 10 on Page 1. 183-13 states that "The (source term) data in this section may not be applicable to cores containing mixed oxide (MOX) fuel." Also, section 3.5.1.1 of the submittal refers to additional fission gas release from MOX fuel. Please provide a scrutable justification for the continued use of the gap fractions in Table 3 with the proposed MOX LTAs.
Best-estimate data developed for use in environmental assessments are generally not acceptable in design basis safety evaluations.
Response
Release fractions are provided in various technical documents (References Q3(g)-1, Q3(g)-2, and other regulatory guides). These need to be applied based upon the licensing basis of the station. As described in response to Radiological Consequences Question 3(i), the licensing basis for Catawba includes both "new" source terms (AST, Reference Q3(g)-l) and "classical" source terms (TID, Reference Q3(g)-2) since the station is in the process of transitioning from the latter to the former. With both being applicable to Catawba, and both historically used for low enrichment uranium (LEU) fuel, both types must be evaluated for validity to mixed oxide fuel (MOX) releases. The support for applicability of these release fractions to MOX fuel originates from the work of an expert panel in both cases.
Table 3 of Reference Q3(g)-1 provides release fractions for non-LOCA fission product gap inventory. Reference Q3(g)-l was based upon work done by an expert panel and described in Reference Q3(g)-4. A similar expert panel was convened to follow-up on this work specifically by assessing "the applicability of NUREG-1465, and if possible, to define a revised accident source term for regulatory applications to reactors using high bumup... low enriched uranium (LEU) fuel and to reactors using mixed oxide (MOX) fuel." The results of this expert panel were published in Reference Q3(g)-3. This work and report were contracted to Energy Research, Inc. and performed under the auspices of the Nuclear Regulatory Commission.
This panel reviewed Reference Q3(g)-4 and provided their opinion on revised LEU release fractions for high burnup fuel. These fractions reflected changes in opinions and the review of more recent data. Table 3.1 of that document displays the updated opinions of this panel and the current guidance from Reference Q3(g)-4. While the opinions were not unanimous in all cases, for high burnup LEU fuel the benefit of further data resulted in an increase in the Tellurium group from 0% to 0.5% release and in noble gases from 5% to 7%. All other release fractions remained unchanged for the gap release.
The panel's opinions on MOX fuel release fractions are presented in Table 3.12 of Reference Q3(g)-3. These gap release fractions reflect the known data up to the time of release of that report. The panel was not in unanimous agreement on the release fractions, but the preponderance of the opinions would retain the gap release fractions for 105
halogens, noble gases and alkali metals at 5% as they are in Reference Q3(g)-4 (note that one panel member was of the opinion that these should be raised to 7%). The opinion on the Tellerium group was split with three of the five panel memibers in favor of no change (0% release fraction) while the other two were in favor of the same release fraction for this group as was updated for the high burnup LEU release fractions (0.5%).
In the case of each isotopic group, the opinion of each member of the expert panel was that the MOX fuel release fraction was of the same value or less than the value which would be proposed if Reference Q3(g)-4 were updated. This indicates that the expert panel saw much similarity between the expected gap release rates of LEU and MOX fuel.
Since Reference Q3(g)-l, Table 3 is based upon expert panel work which was published in Reference Q3(g)-4 and the panel saw similarities in gap release rates between LEU and MOX fuel, it could be inferred that the gap release rates in Reference Q3(g)-1, Table 3 should also be valid for MOX fuel gap releases. The applicability of this table to higher bumup fuel was examined. This evaluation was based upon third cycle peaking factors for both LEU and MOX fuel lead assemblies. The linear heat rate requirement for application of these release fractions at higher burnups was satisfied for both types of fuel.
Nevertheless, it is recognized that current data comparisons show fission product gas release from MOX fuel pellets is generally higher than fission product gas release from LEU fuel (Chapter 5 of Reference Q3(g)-5). Figure Q3(g)-1 shows Framatome ANP data on fission gas release for European MOX fuel and LEU fuel. Although there is some data scatter for the MOX and LEU gas releases (as can be seen by the range of releases for a given bumup) this figure indicates higher fission product gas release for MOX fuel.
To account for this performance, sensitivity studies were performed on the assumed gap fractions to study the impact on dose analyses. The release fraction of all halogens and noble gases (as provided in Table 3 of Reference Q3(g)-l) were increased by 50% to provide a safety case for the MOX lead fuel assemblies. The alkali metal release fraction was not changed. The various gap release fractions are summarized in Table Q3(g)-1.
This rationale is also extended to the TID source term analyses that were evaluated.
Since AST release fractions were increased in the FHA analyses to account for the increase in gas in the fuel-cladding gap and to provide a bounding safety case for MOX fuel lead assemblies, the same level of conservatism was applied to the TID-based accidents that were evaluated. Therefore, these evaluations included a factor of 1.5 to mimic an increase in the TID release fractions as was done for the AST release based analyses.
As discussed in the response to Radiological Consequences Questions 3(a) and 3(f), the analysis work that was performed for the fuel handling and weir gate drop accidents showed that the TEDE dose is driven by the I-131 activity release. Therefore, the increase in the I-131 of 50% resulted in a comparable increase in doses (relative to the same fuel type with nominal release fractions, see response to Radiological Consequences Question 3(b)). With the elevated release fractions, the dose results were 106
still acceptable when compared to limits established and discussed in response to Reactor Systems Question 12.
See responses to Reactor Systems Question 12 and Radiological Consequences Questions 3(a), 3(b), 3(f) and 3(j) for further information.
References Q3(g)-l USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q3(g)-2 USAEC, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.
Q3(g)-3 ERI(NRC 02-202, "Accident Source Tenn for Light-Water Nuclear Power Plants: High Bumup and Mixed Oxide Fuels," November 2002.
Q3(g)-4 USNRC, NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," February 1995.
Q3(g)-5 IAEA Technical Report Series No. 415, "Status and Advances in MOX Fuel Technology," International Atomic Energy Agency, Vienna, Austria, 2003.
Table Q3(g)-1 Various Non-LOCA Fission Product Gap Release Fractions ERIIRC 0-202 ERIINRC 02-202 Reg Guide Raised Isotope Group NUREG 14651 (high burnup 1MOX3fue' 3 5 1.1334 AsT6 1-131 5%
5%
5%
8%
12%
Kr-85 5%
7%
5%
10%
15%
Other Noble 5%
7%
5%
5%
7.5%
Gases Other Halogens 75 5%
5%%
5 7.5%
Alkali Metals 5%
5%
5%
12%
12%
1Reference 4 2 Reference 3, Table 3.1 for high bumup fuel 3 Reference 3, Table 3.12 for MOX fuel 4 Reference 1 5 Since the opinion of the panel was not unanimous, these values represent the majority of the panel's opinion.
6 Values used In the sensitivity analyses for the bounding safety case for MOX lead assemblies. These represent a 50% Increase over Reference 1 with the exception of alkali metals.
107
Figure Q3(g)-1 Fission Gas Release Data for European MOX and LEU Fuels 7
(J5 75;5 ;<¶.
O 3
~
2-0 10000 20000 30000 40000 50000 60000 70000 80000 Bumup MWd/MTHM 108
Question 3:
- h. The CLB fuel handling accident at McGuire is based on the guidance of Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors." Footnote 1 on page 25.2 of the guide states that the assumptions given in the guide are valid only for oxide fuels of the types currently in use. Also, Section 3.5.1.1 of your submittal refers to additional fission gas release from MOX fuel. Please provide a scrutable justification for the continued use of the gap fractions specified in Regulatory Position C.2.d of Regulatory Guide 1.25 with the proposed MOX LTAs. Best-estimate data developed for use in environmental assessments are generally not acceptable in design basis safety evaluations.
Response
Because the MOX fuel lead assembly license amendment request has been amended to apply to Catawba only, no response is provided to this question.
Question 3:
- i. Section 3.5.1.1 of the submittal refers to additional fission gas release from MOX fuel.
Please explain the impact of this increased fission gas production and associated increase in fuel pin pressures on the iodine decontamination factors provided in Regulatory Guides 1.25 and 1.183 that are predicated on gas pressures less than 1200 psig.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response does not address McGuire.
The effect of an increase in spent fuel pin pressures (due to increased gas production in a confined space) is to decrease the decontamination factor (DF) of the water that the isotopes are traveling through. Thus, a greater percentage of the source term (and more curies of activity) is released as the pin pressure increases.
The allowed spent fuel pool pin pressure for Catawba is 1300 psig. This value is based upon analyses using Reference Q3(i)-2 methodology. Using this methodology, the calculated value of the spent fuel pool (with 23 feet of water) elemental DF at 1200 psig is 580. At 1300 psig, it is calculated to be 542. Reference Q3(i)-1 specifies that an overall DF of 200 should be used in the analysis. Given an organic DF of 1 from Reference Q3(i)-l, a corresponding elemental DF of 350 is derived for use in the fuel handling accident (FHA) and weir gate drop (WGD) analyses.
This elemental DF value is less than the calculated value of 542 for 1300 psig spent fuel pool pin pressure, and therefore provides additional conservatism.
An analysis of internal fuel pin pressures in the spent fuel pool has been performed. This analysis calculated that the pin pressures of the MOX fuel assemblies would remain below 1300 psig. The DFs used in the MOX analyses meet the Reference Q3(i)-1 (Catawba current licensing basis for FHAs) requirement for an overall DF of 200 and prvide substantial conservatism when compared to those values calculated for 1300 psig pin pressures.
See response to Radiological Consequences Question 3(a) for further information.
109
References Q3(i)-1 USNRC Regu!a'o-y Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q3(i)-2 Westinghouse, WCAP 7828, "Radiological Consequences of a Fuel Handling Accident," December 1971.
Question 3:
- j.
The licensee's November 25, 2002, submittal on a full scope alternative source term (AST) at Catawba, currently under NRC staff review, does not address the proposed MOX LTAs. The MOX LTA amendment analyses do not address AST and TEDE. Please explain how these differences are planned to be resolved, and comparable differences for any other proposed license amendments for Catawba or McGuire that are currently pending.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, this response does not address McGuire.
The use of MOX fuel and Alternative Source Term (AST, Reference Q3(j)-l) are two separate issues. The Catawba AST fuel handling accident (FHA) analysis was reviewed and approved in April 2002. Catawba currently has a license amendment request (LAR) under review by the NRC Staff for full adoption of AST through approval of a loss of coolant accident (LOCA) analysis using AST. The LOCA analysis (Reference Q3(G)-3) was submitted in November 2002.
Its review has not been completed. These two accident analyses must be reviewed and approved through a Safety Evaluation Report (SER) prior to the full adoption of AST. Analyses of other accident scenarios are being performed in anticipation of the eventual adoption of AST for Catawba as the opportunity arises. The steam generator tube rupture and the main steamline break accident AST analyses are in the process of being developed.
Catawba MOX fuel handling accident analyses are performed in accordance with AST and reported in terms of TEDE, because Catawba has approval to use AST for the FHA. All other analysis and evaluations in support of the four MOX lead assemblies are performed in accordance with the classical source term (TID, Reference Q3(6)-2) and reported in terms of thyroid and whole body. Future submittals for Catawba may reflect a different licensing basis as it continues transitioning from TID to AST.
See the responses to radiological consequences questions 3(a) and 3(e) for further information.
References Q3(j)-1 USNRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q3(j)-2 USAEC, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.
Q3(j)-3 Letter from G. R. Peterson (DPC) to USNRC, "Proposed Technical Specifications and Bases Amendment: Technical Specification and Bases 3.6.10 Annulus Ventilation System, Technical Specification and Bases 3.6.16 Reactor Building, Technical Specification and Bases 3.7.10 Control Room Area Ventilation System (CRAVS), Technical Specification and Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), Technical Specification and Bases 3.7.13 Fuel Handling Ventilation Exhaust System (FHVES), Technical Specification and 110
i Bases 3.9.3 Containment Penetrations, Technical Specification and Bases 5.5.1 Ventilation Filter Testing Program (VFTP)," November 25, 2002.
- 4. The NRC staff has a concern regarding whether the radiological analyses described in the submittal of February 27,-2003, are consistent with the respective licensing bases of Catawba and McGuire. This is a concern since the licensee has stated that the results for Catawba provide a basis for concluding that the doses at McGuire would also be acceptable. Please resolve the apparent discrepancies discussed below in a manner that is consistent with the licensing bases at both sites and that provides a clear understanding of what the licensing basis will be following implementation of this amendment.
- a. On Page 3-34 of Attachment 3 of the submittal, the first paragraph states that:
The analyses were conducted in accordance with the regulatory positions of Regulatory Guide 1.25 and the guidelines in Standard Review Plans (SRPs) 15.7.4, 9.4.1, and 9.4.2.
By letter dated December 20, 2001, the licensee requested an amendment for Catawba that would selectively replace the TID14844 source term used in the fuel handling and weir gate accidents with an alternative source term and replace the previous whole body, skin, and thyroid doses with TEDE. As such, the current licensing basis at Catawba includes AST and TEDE and RG 1.183 (as they apply to fuel handling and weir gate accidents). However, the analyses results in this section are reported in terms of whole body, skin, and thyroid, which, while consistent with the licensing basis for McGuire, are inconsistent with the licensing basis for Catawba.
Response
As discussed in the response to Radiological Consequences Question 3(j), the Catawba source term licensing basis is in transition. Therefore, the evaluations and analyses performed for the MOX fuel lead assembly program were done in accordance with Catawba's current licensing basis. Accordingly, the fuel handling and weir gate drop analyses were performed in accordance with alternative source term (AST, Reference Q4(a)-1) and reported in terms of TEDE, and all others were performed in accordance with classical source term releases (TID, Reference Q4(a)-
- 2) and reported in terms of whole body and thyroid.
See the response to Radiological Consequences Questions 3(e) and 3(j) for further information.
References Q4(a)-l USNRC Regulatory Guide 1.183, "Altemative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
Q4(a)-2 USAEC, Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.
Question 4.
- b. Similarly, the dose results for the new accident analyses addressed in section 3.7.3.5 are expressed in terms of TEDE. The licensee has not requested the use of AST and TEDE for McGuire pursuant to 10 CFR 50.67. The licensee re-analyzed the postulated fuel handling and weir gate events in support of its December 20, 2001, application for changes to certain refueling mode technical specifications at Catawba, based on a selective implementation of 111
the AST. That application was the subject of amendment numbers 198 and 191 to the Catawba, Units 1 and 2 operating licenses, as issued on April 23, 2002.
The NRC staffs position is that, unless it is authorized pursuant to 10 CFR 50.67, the TEDE dose quantity cannot be used for demonstrating compliance with Part 50 requirements.
Response
The MOX fuel lead assembly license amendment request has been amended to apply to Catawba only. Therefore, no response is provided to this question.
112
ENVIRONMENTAL IMPACTS QUESTION In sections 5.6.1, "Plants Effluents," and 5.6.2, "Impacts to Human Health," it is concluded that there are "... no anticipated changes in the types or amounts of plant effluents resulting from the use of the four MOX fuel lead assemblies," and that "Plant releases will comply with all regulatory limits with MOX fuel lead assemblies in the reactor, such that there will be no impact on public health and safety."
Please provide a technical basis to support the conclusion that your plant process systems can handle the MOX fuel and that radiological effluents will remain within regulatory limits.
Response
The conclusion that there are no anticipated changes in the type or amounts of plant effluents resulting from the use of MOX fuel lead assemblies is based on the similarity of MOX fuel to current LEU fuel, both from a fuel design and a fission product inventory perspective, and on the fact that only four out of 193 fuel assemblies will contain MOX fuel.
Fuel reliability for the Mark-BW/MOX1 fuel design is expected to be similar to the reliability of advanced Mark-BW low enriched uranium (LEU) fuel (Reference E-1, Section 7.1.3). European experience with MIMAS-produced MOX fuel rods is comparable to the operating experience of LEU fuel. Of the fuel rod failures observed in European reactors, none have been attributed to the use of MOX fuel (Reference E-1, Section 7.2.4). Thus, the conclusion is that there should be no significant difference in the reliability of MOX fuel from that of LEU fuel; i.e., the MOX fuel rod failure rates should be similar to that of LEU fuel rods. European MIMAS-produced MOX fuel assemblies have experienced [
], or [
] failure rate per assembly. At this failure rate the probability of any failure in the four MOX fuel lead assemblies is low, on the order of[
].
With similar fuel reliability, the other factor that affects radiological effluents is the isotopics of the irradiated fuel. As fuel is irradiated, both activation and fission products are created.
The activation products are created in the reactor coolant and fission products are produced inside the fuel rods. Activation products that are created are a function of (i) impurities in and chemistry of the reactor coolant and (ii) the thermal neutron flux that the materials see.
Impurities and chemistry are independent of the fuel type. Thermal flux is significantly lower in MOX fuel than in LEU fuel, which would tend to lower activation products.
However, for four lead assemblies this is expected to be an insignificant effect.
Fission product inventories and gap inventories in particular are of the same order of magnitude in both MOX fuel and LEU fuels. In particular, the amount of iodine and noble gas that would be released into the reactor coolant in the event of a leaking fuel rod would be similar. This is evident by examining Table E-1, which compares the fission product inventories for MOX fuel and LEU fuel with equivalent initial reactivity. Additionally, any liquid or gaseous effluents would be processed by the plant liquid waste and waste gas systems prior to release. The plant treatment systems are fully capable of treating these effluents since the types of radiological isotopes in MOX and LEU fuel are the same and the curie content of MOX fuel is of the same order of magnitude as LEU fuel. This leads to the conclusion that there will be no impact on the public health and safety and that effluent releases will remain within regulatory limits.
113
This conclusion is supported by experimental and operational experience with MOX fuel in Europe. In the 1989 EDITHMOX 01 experiment in the SILOE research reactor, radionuclide release rates from a failed MOX fuel rod were measured during irradiation and found to be comparable to release rates from a failed LEU fuel rod (Reference E-2). This was followed by the 1997 EDITHMOX 02 experiment, also performed at the SILOE reactor but this time involving a high burnup (56 GWd/Mthm) MOX fuel rod. The release rates at EDITHMOX 02 were also comparable to LEU fuel, and less than or equal to the values obtained at EDITHMOX 01 (Reference E-3). Finally, in 1993-1994 a MOX fuel assembly operated with a failed fuel rod for parts of two cycles at the Dampierre 1 power plant. EDF and CEA conducted a special monitoring program during the second cycle. The measurements indicated that the release reates of gaseous fission products were similar to those observed with defective LEU fuel (Reference E-4).
In summary, the bases for the conclusion that plant process systems can handle the MOX fuel and that radiological effluents will remain with regulatory limits are:
- 1. MOX fuel-reliability should be similar to LEU fuel, and no MOX fuel failures are anticipated. Without MOX fuel failures, the MOX fuel assemblies should not release fission products that impact plant systems and lead to radioactive effluents.
- 2. MOX fuel fission products are similar to LEU fuel, and activation products should be lower. Therefore, in the unlikely event that a MOX fuel rod should fail, the impact on plant systems and efflu-ents will be comparable to a LEU fuel failure.
- 3. European experimental data and plant operating data support point (2), above.
References E-1. BAW-10238P Revision I, MOXFuel Design Report, Framatome ANP, May 2003.
E-2. D. Parrat, Y. Musante, A. Brissaud, "Mixed Oxide Fuel in Defective Experimental Rod EDITHMOX 1: Irradiation Results and Metallographic PIE," International Atomic Energy Agency Technical Document 709, 1992.
E-3 D. Parrat, A. Harrer, "Failed High Bum-up MOX Fuel Performance: The EDITHMOX 02 Analytical Irradiation," American Nuclear Society Light Water Reactor Fuel Performance Meeting, Park City, Utah, April 10-13, 2000.
E-4 D. Parrat, C. Leuthrot, A. Harrer, D. Dangouleme, "Behaviour of a Defective MOX Fuel Rod in a PWR," International Atomic Energy Agency Technical Document 941, 1995.
114
Table E-1 Comparison of Typical Isotopic Inventory for MOX Fuel and LEU Fuel I f
'Diff~~~~~~~~~~~~erence i%)
stE 2Curie LEU).
es(MOX)
(Mo LEU)/LEU 1130 2.09E+04 1.94E+04
-7.2%
1131 7.99E+05 8.30E+05 3.9%
1132 1.17E+06 1.20E+06 2.6%
1133 1.63E+06 1.61 E+06
-1.2%
1134 1.79E+06 1.73E+06
-3.4%
1135 1.56E+06 1.56E+06 0.0%
Kr 83m 9.02E+04 6.90E+04
-23.5%
Kr 85m 1.81 E+05 1.24E+05
-31.5%
Kr 85 5.31 E+03 3.16E+03
-40.5%
Kr 87 3.61 E+05 2.41 E+05
-33.2%
Kr 88 4.98E+05 3.22E+05
-35.3%
Kr 89 6.11 E+05 3.68E+05
-39.8%
Xe131m 1.11 E+04 1.1 8E+04 6.3%
Xe133m 4.92E+04 5.03E+04 2.2%
Xe133 1.56E+06 1.55E+06
-0.6%
Xe135m 3.48E+05 3.71 E+05 6.6%
Xe135 2.69E+05 5.29E+05 96.7%
Xe137 1.43E+06 1.47E+06
-1.3%
Xel 38 1.36E+06 1.28E+06
-5.9%
Br 83 8.93E+04 6.83E+04
-23.5%
Br 85 1.80E+05 1.24E+05
-31.1%
Br 87 2.84E+05 1.84E+05
-35.2%
Cs134 1.03E+05 9.97E+04
-3.2%
Cs136 3.63E+04 5.99E+04 65.0%
Cs137 6.22E+04 6.39E+04 2.7%
Cs138 1.49E+06 1.42E+06
-4.7%
Cs139 1.38E+06 1.30E+06
-5.8%
LEU Fuel U-235 Enrichment - 4.0%,
Bumup - 41.3 GWd/MThm MOX Fuel Pu Concentration - 5.0%
Bumup - 41.3 GWd/MThm 115