ML20195H042
| ML20195H042 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 11/17/1998 |
| From: | Tam P NRC (Affiliation Not Assigned) |
| To: | NRC |
| References | |
| TAC-MA2359, TAC-MA2361, TAC-MA2411, TAC-MA2412, NUDOCS 9811230228 | |
| Download: ML20195H042 (6) | |
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WASHINGTON, D.C. 206MKX101
'+9 * * * * *,o November 17,1998 MEMORANDUM FOR:
Docket File" FROM:
Peter S. Tam, Senior Project Manager Project Directorate 11-2
(
Division of Reactor Projects - 1/11
).
m Office of Nuclear Reactor Regulation
SUBJECT:
CATAWBA AND MCGUIRE NUCLEAR STATIONS -
ELECTRONIC COMMUNICATION, QUESTIONS ON THE LICENSEE'S REQUEST FOR AMENDMENT DATED JULY 22, 1998 (TAC MA2359, MA2361, MA2411, MA2412 )
The attached questions were transmitted by e-mail today to Steve Warren of Duke Energy Corporation (DEC) to prepare him and others for an upcoming telephone call. This memorandum and the attachment do not convey a formal request for information or represent an NRC staff position.
Docket Numbers 50-369,50-370, 50-413 and 50-414 Distribution PUBLIC H. Berkow Y
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9811230228 981117 PDR ADOCM 05000369 eg i}-Q P
REQUEST FOR ADDITIONAL INFORMATION DPC-NE-2009. "DPC WESTINGHOUSE FUEL TRANSITION REPORT"
(
Reference:
Letter, M. S. Tuckman to NRC, July 22,1998) 1.
Section 3.2 of DPC-NE-2009 states that conceptual transition core designs using the RFA design have been evaluated and show that current reload limits remain bounding with respect to key physics parameters, and that in the event that one of the key parameters is exceeded, the evaluation process described in DPC-NE-3001-PA would be performed.
(a)
Describe the evaluation and the result of the conceptual transition core design.
(b)
From the statement it appears that the evaluation process described in DPC-NE-3001-PA will not be performed unless one of the key parameters is exceeded.
Without actual analysis of the RFA transitional or full cores, how is it determined that any of the key parameters is exceeded?
2.
To demonstrate that the currently approved CASMO-3/ SIMULATE-3P methods and nuclear uncertainties in DPC-NE-1004-PA are applicable to the RFA design, Section 3.2 cites the analyses performed using Sequoyah Unit 2 Cycles 5,6 and 7, as well as a 10 CFR 50.59 USQ evaluation. It is stated that the Sequoyah cores were chosen because they are similar to McGuire and Catawba and contained both IFBA and Wet Annular Burnable Absorber fuel. Table 3-1 provides the statistical analysis results of nuclear uncertainty factors, which show they are bounded by the uncertainty factors of DPC-NE-1004A.
(a)
Describe any difference between the McGuire/ Catawba RFA cores and the Sequoyah cores analyzed. Describe why these differences would not affect the applicability of the analyses of the Sequoyah cores to McGuire and Catawba.
(b)
Provide the comparison of the analysis results with measured data of boron concentrations, rod worths, and isothermal temperature coefficients.
(c)
Describe the details and results of the 10 CFR 50.59 USQ evaluation 3.
Section 3.2 states that (1) in all nuclear design analysis, both the RFA and the Mark-BW fuel are explicitly modeled in the transition cores, and (2) when establishing Operating and RPS limits (i.e., LOCA kw/ft, DNB, CFM, transient strain), the fuel specific limits or a conservative overlay of the limits are used.
Please elaborate on the mixed core model for nuclear design analyses, and how fuel-specific limits are used.
4.
Section 5.2 states that in using the VIPRE-01 code for the reactor core thermal-hydraulic analysis, the reference power distribution based on a 1.60 peak pin from DPC-NE-2004P-A, Rev.1, was used.
(a)
The report states that this reference pin power distribution "was" used. Will it be used for future RFA reload analyses?
2 (b)
Does the reference pin power distribution used in the core thermal-hydraulic analyses bound all power distribution for the RFA cores for future reload cycles?
5.
Section 5.2 states that in the thermal-hydraulic analysis of the RFA design using VIPRE-01, the two-phase flow correlations will be changed from the Levy subcooled void correlation and the Zuber-Findlay bulk void correlation to the EPRI subcooled and bulk void correlations, respectively. While the sensitivity study provided in the report shows a minimal difference of 0.1% between the minimum DNBRs of 51 RFA CHF test data points calculated with both set of correlations, it was stated in DPC-NE-2004 that the Levy /Zuber-Findlay combination compared most favorably with the Mark-BW test results as the DNBRs of the tests calculated with this combination yielded conservative results relative to the EPRI correlations.
(a)
Discuss whether the EPRI correlations will be used for the RFA design only, or they will also be used for the Mark-BW design.
(b)
If the EPRI correlations will also be used for Mark-BW design, provide justification for their use.
(c)
If the Levy /Zuber-Findlay correlations will continue to be used for Mark-BW fuel design, discuss how the VIPRE-01 code will be used to analyze transient mixed cores having both Mark-BW and RFA fuel designs.
6.
Section 5.7 describes the use of a transition 8-channel RFA/ Mark-BW core model to determine the impact of the geometric and hydraulic differences between the resident Mark-BW fuel and the RFA design, and determine a conservative DNBR penalty to be applied for the transition cores. Table 5-4 presented the statistical DNBRs for the 500 and 5000 case runs for various statepoints including the transition core case of the most limiting statepoint 12. The statistical design limit is chosen to bound both the full RFA cores and RFA/ Mark-BW transition cores for the 5000 case runs.
(a)
Why is the statistical design limit value proprietary information?
(b)
With respect to the statistical core design methodology, describe how the uncertainties of the CHF correlation and the VIPRE code /model are propagated with the uncertainties of the selected parameters of each statepoint for the calculation of the statistical DNBR for each statepoint in Table 5-4.
(c)
With the statistical design limit specified in Section 5.7, is it your intention to use a full core of RFA in the thermal hydraulic analysis for the transition core without the transition core DNBR penalty factor?
7.
Section 2.0 states that the RFA is designed to be mechanically and hydraulically compatible with the Mark-BW fuel. Table 2.1 provides a comparison of the basic design parameters of the two fuel designs, but does not provide a comparison of the hydraulic characteristics of spacer grids. Section 5.2 states that the VIPRE-01 core thermal-hydraulic analyses were performed with applicable form loss coefficients as per the vendor. Table 5.1 provides general RFA fuel specifications and characteristics without the hydraulic characteristics of the spacer grids.
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i 3
(a)
Provide comparisons for the thickness, height, and form loss coefficients of the RFA and Mark-BW fuel spacer grids, including mixing-vane and non-mixing vane structural grids, and intermediate flow mixing grids.
(b)
Provide the form loss coefficients of the spacer grids used in the analyses and in the RFA CHF test assemblies if they are different from the values described in item (a) above.
(c)
Describe the procedures to ensure that the form loss coefficients of the RFA grids are comparable to those used in the SCD analysis and the CHF tests so that both -
the WRB-2M CHF correlation DNBR limit and the statistical core design limit are valid.
8.
Section 6.1.3 states that the thermal-hydraulic methodology described in DPC-NE-3000-PA, Rev.1, with a simplified core model will be used for thermal-hydraulic analysis of UFSAR Chapter 15 Non-LOCA transients and accidents for the RFA design. It also states that (1) no transition core transient analyses are performed as the results determined in Chapter 5 also apply for transient analyses, (2) the simplified core model of DPC-NE-3000-PA used for transient analyses was originally developed with additional conservatism over the 8 channel model used for steady-state analyses to specifically minimize the impact of changes in core reload design methods or fuel assembly design, and (3) should it be determined in the future that transition core transient analyses are warranted, they will be performed accordingly.
(a)
Explain what additional conservatism is provided in using the simplified core model of DPC-NE-3000-PA.
(b)
What is the criterion or criteria used to determine if transition core transient analyses are warranted? How would it be determined that the criteria have been exceeded without RFA transition core analyses?
9.
Regarding rod ejection analysis using SIMULATE-3K, it is stated in Section 6.6.2.2.1 that the transient response is made more conservative by increasing the fission cross sections in the ejected rod location and in each assembly and by applying " factors of conservatism" in the moderator temperature coefficient, control rod worths for withdrawal and insertion, Doppler temperature coefficient, effective delay neutron fraction, and ejected rod worth, etc.
(a)
What are the values of the multiplication factors used for fission cross sections, and how are they determined?
(b)
How are the input multipliers " VAL" in Equations 6.1 and 6.2 determined? Does
" VAL" have a different value for different parameters, such as MTC or DTC? What are the values for these VALs?
(c)
In Equation 6.1, the X's are described as " moderator temperatures." Should they be moderator temperature coefficients?
10.
Regarding the SIMULATE-3K code, there is an optional " frequency transform" approach, under the " Temporal Integration Models," that can be chosen to
4 separate the fluxes into exponential time varying and predominately spatial components, thus accelerating convergence of the transient neutronic solution and preserving accuracy on a coarser time mesh (see Page 5, Ref. 6-9).
(a)
What determines when the " frequency transform" approach should be used?
(b)
What are the consequences of exercising (or not exercising) this option? Please provide technicaljustification and comparisons of results.
l 11.
The licensing analyses of reload cores with the RFA design will use the methodologies described in various topical reports and revisions for the analyses of fuel design, core reload design, physics, thermal-hydraulics, and transients and accidents, which were approved by NRC for analyses of current McGuire/ Catawba cores not having the RFA design. For example, DPC-NE-1004A, DPC-NE-2011-PA, DPC-NF-2010A, and DPC-NE-3001-PA are used for the nuclear design calculations. DPC-NE-2004-PA, DPC-NE-2005-PA, and the VIPRE-01 code are used for the core thermal-hydraulic analyses and statistical core design. DPC-NE-3000-PA, DPC-NE-3001-PA, DPC-NE-3002-A, and RETRAN-02 code are used for non-LOCA transient and accident analyses. Westinghouse small-and large-break LOCA evaluation models described in WCAP-10054-P-A and WCAP-10266-P-A, and related topical reports, are used for the small-and large-break LOCA ana'yses.
Some of these methodologies have inherent limitations, and some have conditions or limitations imposed by the NRC SERs in their applications. Provide a list of the inherent limitations, conditions, or restrictions applicable to the RFA core design from all the methodologies to be used for the RFA reload design analyses, and describe the resolutions of these limitations, conditions and restrictions in the applications to the RFA cores and the transitional RFA/ Mark-BW cores.
12.
Section 8.0 states that TS Figure 2.1.1-1 for the reactor core safety limits will be modified by deleting the 2455 psia safety limit line and making the 2400 psia safety limit line as the upper bound pressure allowed for power operation. Since the upper range of applicability of the WRB-2M CHF correlation for the RFA design is 2425 psia, the 2400 psia safety limit line is within the range of the CHF correlations for the Mark-BW and RFA fuel designs.
However, the safety limit lines in Figure 2.1.1-1 were based on the CHF correlation for the Mark-BW fuel design, in addition to the hot leg boiling limit. Has an analysis
- been performed to ensure these safety limit lines bound the safety limit for the DNBR limit of the WRB-2M correlation for the RFA design?
13.
TS surveillance requirements 3.2.1.2,3.2.1.3, and 3.2.2.2, respectively, require the heat flux hot channel factor F, (x,y,z) and the enthalpy rise hot channel factor F.
(x,y) to be measured periodically using the incore detector system to ensure the values of the total peaking factor and the enthalpy rise factor assumed in the accident analyses and the reactor protection system limits are not violated. To avoid the possibility that these hot channel factors may increase beyond their allowable limits between surveillances, these SRs currently specify a penalty factor of 1.02 for the heat flux and enthalpy rise hot channel factors if the margin to the F, (x,y,z) or F (x,y) has decreased since the previous surveillance. For the reactor core containing the RFA fuel desi n with integral burnable absorbers, larger penalty 0
may be required over certain burnup ranges early in the cycle due to the rate of
(
5 burnout of this poison. Section 8.1 proposes to remove the 2% penalty value from these surveillance requirements and replace them with tables of penalty values as functions of burnup in the Core Operating Limits Report (COLR) to facilitate cycle specific updates. Tables 8-1 and 8-2, respectively, provide " typical values" for the burnup-dependent margin-decrease penalty factors for the heat flux and enthalpy l
rise hot channel factors.
(a)
Provide the actual values of the margin-decrease penalty factors, as well as the l
bases for these values.
(b)
Provide references for the approved methodologies used to calculate these values, and to be included in TS 5.6.5 as a part of acceptability for COLR.
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