ML20206S334

Forwards Proprietary & non-proprietary Responses to NRC 981209 & 990105 RAIs Re Util Lars,Permitting Use of W Fuel at McGuire & Catawba Stations.Proprietary Info Withheld,Per 10CFR2.790

ML20206S334
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	01/28/1999
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20136E397	List: ML20206S334
References
NUDOCS 9902040173
Download: ML20206S334 (39)

Text

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gg Duka Energy C:rpor;ti:n gg "

526 South Church Street I!O. Box 1006 (EC07H)

Charlotte, NC 28201-1006 M. S. Tuckman (704) 382-2200 omCF Executive Vice hesident (704) 382-4360 ux

Nuclear Generation January 28, 1999 U. S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 ATTENTION

Document Control Desk 4

Subject:

Duke Energy Corporation McGuire Nuclear Station Units 1 & 2 ,

Docket Nos. 50-369, 50-370 l

Catawba Nuclear Station Units 1 & 2 Docket Nos. 50-413, 50-414 Response to NRC Requests for Additional Information l on License Amendment Requests for McGuire and l Catawba Nuclear Stations {

l This submittal contains information that Duke Energy l Corporation considers PROPRIETARY and is being made pursuant '

to 10CFR 2.790.

By letters dated December 9, 1998 and January 5, 1999 the NRC requested additional information on Duke Energy Corporation's July 22, 1998 license amendment requests (LARs) for the McGuire Nuclear Station, Units 1 & 2; and the Catawba Nuclear Station, Units 1 & 2 Technical Specifications. These LARs would permit use of Westinghouse fuel at McGuire and Catawba.

Topical Report DPC-NE-2000P/DPC-NE-2009 was also included in the July 22, 1998 Duke submittal.

The thirteen questions contained in the December 9, 1998 NRC letter, and the corresponding Duke answers, are provided in the attachments to this letter. A proprietary version and a j non-proprietary version of the Duke response are attached to this letter.

i (20 Some of the information contained in Attachment 1 is \

considered proprietary. In accordance with 10CFR 2.790, Duke Energy Corporation requests that this information be withheld from public disclosure. An affidavit which attests to the C luu }cr R E ri c L 173 990128 h20 POCK P 05000369 PDR-

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U. S. Nuclear Regulatory Commission January 28, 1999 Page 2 proprietary nature of the affected information is included with this letter. A non-proprietary version of the Duke response is included as Attachment 2 to this letter.

Please address any comments or questions regarding this matter to J. S. Warren at (704) 382-4986.

Very truly yours, j

. b* a -

M. S. Tuckman Attachments xc (w/o Attachment 1):

Mr. L. A. Reyes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. F. Rinaldi, Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-14H25 Washington, D. C. 20555-0001 Mr. P. S. Tam, Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-14 H25 Washington, D. C. 20555-0001 Mr. S. M. Shaeffer NRC Senior Resident Inspector McGuire Nuclear Station Mr. D. J. Roberts NRC Senior Resident Inspector Catawba Nuclear Station l

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1 l U. S. Nuclear Regulatory Commission

January 28, 1999 Page 3 1

AFFIDAVIT

1. I am Executive Vice President of Duke Energy Corporation; and as such have the responsibility for reviewing information sought to be withheld from public disclosure in connection with nuclear power plant licensing; and am authorized on the part of said Corporation (Duke) to apply for this withholding.

2. I am making this affidavit in conformance with the provisions of 10CFR 2.790 of the regulations of the

'Juclear Regulatory Commission (NRC) and in conjunction with Duke's application for withholding, which accompanies this affidavit.

3. I have knowledge of the criteria used by Duke in i designating information as proprietary or confidential. l i

4. Pursuant to the provisions of paragraph (b) (4) of 10CFR 2.790, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public l disclosure is owned by Duke and has been held in i confidence by Duke and its consultants.

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(ii) The information is of a type that would customarily be held in confidence by Duke. The information consists of analysis methodology details, analysis results, supporting data, and aspects of development programs relative to a method of analysis that provides a competitive advantage to Duke.

fb M. S. Tuckman f

(Continued) f 1

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U. S. Nuclear Regulatory Commission January 28, 1999 Page 4 (iii)The information was transmitted to the NRC in confidence and under the provisions of 10CFR 2.790, it is to be received in confidence by the NRC.

(iv) The information sought to be protected is not available in public to the best of our knowledge and belief.  !

(v) The proprietary information sought to be withheld in this submittal is that which is marked in the proprietary version of the Duke response to NRC l requests for additional information dated December 9, 1998 and January 5, 1999. The subject of these requests for additional information is a Duke i license amendment request dated July 22, 1998 and accompanying topical report designated DPC-NE-

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2009P, Duke Power Company Westinghouse Fuel Transition Report. The information of concern is omitted from the non-proprietary version of the Duke response. This information enables Duke to:

(a) Respond to Generic Letter 83-11, Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions.

(b) Perform core design, fuel rod design, and thermal-hydraulic analyses for the Westinghouse Robust Fuel Assembly design.

(c) Simulate UFSAR Chapter 15 transients and accidents for McGuire and Catawba Nuclear Stations.

(d) Perform safety evaluations per 10CFR50.59.

(e) Support Facility Operating Licenses / Technical Specifications amendments for McGuire and Catawba Nuclear Stations.

f. b. =::__ ,

M. S. Tuckman (Continued)

U. S. Nuclear Regulatory Commission January 28, 1999 Page 5 (vi) The proprietary information sought to be withheld from public disclosure has substantial commercial j value to Duke.

(a) It allows Duke to reduce vendor and consultant l expenses associated with supporting the operation and licensing of nuclear power plants.

1 (b) Duke intends to sell the information to j nuclear utilities, vendors, and consultants i for the purpose of supporting the operation and licensing of nuclear power plants.

(c) The subject information could only be duplicated by competitors at similar expense ,

to that incurred by Duke.

5. Public disclosure of this information is likely to cause harm to Duke because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring commensurate expense or allowing Duke to recoup a portion of its expenditures or benefit from the sale of the information.

8. Gk M. S. Tuckman (Continued) l

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i U. S. Nuclear Regulatory Commission January 28, 1999 Page 6 M. S. Tuckman, being duly sworn, states that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth within are true and correct to the best of his knowledge.

M. S. Tuckman, Executive Vice President subscribed and sworn to before me this N day of

- A AI(44D , 1999 J .,

1 (1LLf >W Notary I/ubl'ic My Commission Expires: l

.8M 2 2 , 2.00/

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l Attachment 2 l Response to NRC Requests for Additional Information Dated December 9, 1998 and January 5, 1999 Applicable to Duke Energy Corporation License Amendment Requests Dated

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July 22, 1998

• • • • Non-proprietary Version ****

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Attachment 2 (Non-Proprietary)

( I. Section 3.2 of DPC-NE-2009P states that conceptual transition core designs using the Robust Fuel Assembly (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.

I (b) Based on 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?

l Response la:

Conceptual Westinghouse RFA transition core (* gns were setup and evaluated using NRC approved codes and methods. The evaluation performed considered the effects of partial and full RFA cores. The purpose of the evaluation was to determine the acceptability of the current licensing bases transient analyses. Key safety parameters were calculated for the conceptual core designs and compared against reference values assumed in the Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analyses. Examples of some of the key parameters calculated include Doppler temperature coefficients, moderator temperature coefficients, control bank wonh, individual rod worths, boron concentrations, differential boron worths and kinuics data. A summary of the key parameters important to the licensing bases transient analyses are provided in Table 2-1 of DPC-NE-3001. The evaluation demonstrated the expected neutronic j similarities between reactor cores loaded with Westinghouse RFA fuel and with Mk-BW fuel, and the acceptability of key safety parameters assumed in the UFSAR Chapter 15 accident analyses.

Response ib:

Key physics parameters important to the UFSAR Chapter 15 accident analyses are calculated for each reload core using NRC approved methodology to confirm that these parameters are bounded by values assumed in the reference UFSAR Chapter 15 accident evaluations. This check is always performed for each new core design. If the key safety parameters assumed in the reference safety analysis are determined to bound the reload core values, then no additional analyses are required. However, if a key physics parameter is not bounded by the reference value, then the affected accidents will be re-analyzed using the new key physics parameter, or the core will be re-designed to produce an acceptable result.

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Attachment 2 (Non-Proprietary)

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 unreviewed safety question (USQ) evaluation. It is stated that the Sequoyah j l cores were chosen because they are similar to McGuire and Catawba and contained both Integral Fuel Burnable Absorber (IFHA) and Wet Annular Burnable Absorber i fuel. Table 31 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 w hy 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 ef the 10 CFR 50.59 USQ evaluation. l Response la:  !

The primary reason for benchmarking the Sequoyah Unit 2 cores was to confirm the fidelity of the CASMO-3/ SIMULATE-3 code suite for analyzing reactor cores containing integral fuel burnable absorbers (IFBA). While the introduction of the IFBA product is not considered a major j design change, and therefore the benchmarking of this product is not required by the SER l

requirements of DPC-NE-1004-PA, a conservative approach was adopted to perform benchmark l calculations to confirm the acceptability of the current nuclear uncertainty factors. Benchmark l

calculations were performed using measured data from Sequoyah Unit 2 Cycles 5,6 and 7.

The Westinghouse Nuclear Design Reports for Sequoyah Unit 2 Cycles 5,6, and 7, the McGuire  !

and Catawba Updated Final Safety Analysis Report (UFSAR) and Section 2.0 of DPC-NE-2009 j were reviewed to determine the differenen between the Sequoyah cores analyzed in the benchmark calculations, and McGuire/ Catawba RFA core designs. A list of differences is {

provided below.

a. The Sequoyah cores modeled and analyzed in the Sequoyah benchmark calculations contained the Westinghouse Vantage-5H (V5H) fuel design. The V5H fuel design is <

geometrically (ie pellet diameter, gap and clad thickness and assembly envelope) equivalent l to the RFA fuel design to be used in the McGuire and Catawba cores. Differences between the V5H and RFA fuel design are primarily mechanical and, as a result, do not impact the l nuclear characteristics of the fuel assemblies. Specific differences between the V5H and RFA fuel design are summarized below. I e

Zirlo* is used for the fuel rod clad, guide tubes, instrument tubes and mixing grids in the RFA fuel design. The V5H design uses Zr-4 for these components. '

The RFA fuel design has thicker instrument and guide tubes than the V5H design in i order to improve structural stability.

The grid design for the RFA design has been modified (optimized vane angles and window size) to improve thermal performance.

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I Attachment 2 i

(Non Proprietary) e The RFA design Duke intends to use has a pre-oxide coating on the bottom of the fuel rods, longer fuel rod end plugs and a protective bottom grid. The V511 fuel design used at Sequoyah did not have these features.

The RFA design employs intermediate mid span mixing grids. The V511 design used at Sequoyah did not use mid span mixing grids.

Neutronically, Zirlo and Zr-4 are equivalent. The changes in instrument tube and guide tube thickness does not impact core modeling as long as they are accounted for in the generation of cross sections and few group constants. The pre-oxide coating does not impact the modeling of the fuel rod or the neutronic properties of Zirlo The fuel rod end plugs are neutronically unimportant because they are located outside of the active fuel region. The mixing grids are specifically accounted for in the neutronics models, therefore, the use of a 1 modified grid design, the addition of the protective bottom grid and mid span mixing grids l should not impact model performance. In summary, the differences in the RFA and V5H fuel designs are primarily mechanical and do not impact the nuclear performance of the fuel assembly. Design features that do impact the neutronics (ie. mid span mixing grids) are specifically accounted for in the nuclear models. Therefore, the results and conclusions reached based on the analysis of the Sequoyah core designs are applicable to the RFA fuel design.

b. The Sequoyah cores that were benchmarked contained both 1.0x and 1.5x IFBAs with rod patterns containing between 48 and 128 IFBA rods. The IFBA loadings (l.0x and 1.5x) and the number of IFBA rods per assembly are representative of the IFB A loadings and the number of IFB A rods expected to be used in McGuire and Catawba core designs. However, the IFB A rod patterns used in the Sequoyah core designs and the IFBA rod patterns that will be used in the McGuire and Catawba core designs are different. The changes in IFBA rod patterns are the result of Westinghouse optimizations that were performed [

] The optimized IFBA rod patterns will be used in the McGuire and Catawba RFA core designs. In addition, all combinations of IFB A loading and rod patterns are explicitly modeled to account for the impact of any design change in the analysis of each reload core design.

The Sequoyah benchmark calculations that were performed demonstrate the acceptability of the CASMO-3/SIMULATF 3 model to accurately calculate core reactivity, reactivity parameters and power distributions for representative IFBA rod loadings and rod configurations. Changes in the IFB A rod configurations primarily affect intra-assembly peaking and not integral and local nodal power distributions which are the parameters that are j measured. Consequently, the results from the benchmark analysis are not expected to change i as the result of changing the IFB A rod pattern design.

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c. The fuel management strategy (Iow leakage - ring of fire core designs), the number of fuel assemblies in the reactor core and the core power rating are the same between McGuire, Catawba and Sequoyah. Ilowever, there are differences in the reactor coolant flow rate and l

core inlet temperature. The reactor coolant flow rate at Sequoyah is approximately 3.0% less I than at McGuire or Catawba. The core inlet temperature at Sequoyah is ~547 F versus l

~555 F at McGuire and Catawba. Core inlet How and temperature are input variables to the .

nuclear model and are therefore specifically accounted for. As a result, the performance of  !

I the nuclear model and the applicability of the benchmark results are not expected to change  !

due to the aforementioned core inlet How and temperature differences.

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. 4 Attachment 2 (Non Proprietary) l l d. The Control Bank (Bank D) for Sequoyah Unit 2 is comprised of 9 control rods versus 5 control rods for McGuire and Catawba. Since control bank locations are specifically l modeled, and because during normal operation control banks are positioned near all rods out (ARO), the impact of this difference on the results of the benchmark analysis is negligible.

Response 2b:

l Comparisons between Duke predicted and measured zero power physics testing (ZI PT) results are shown below for Sequoyah Unit 2 Cycles 5,6 and 7. The 7PIrr results included comparisons of critical boron concentrations, control rod worths and isothermal temperature coefficients. Excellent agreement between predicted and measured results is generally observed.

The large percent differences between predicted and measured control rod worths for Control Bank A in cycles 5 and 6 is primarily the result of the low worth of these banks and to a lessor extent a slight mis-prediction (~1.0%) in the local power distribution. The observed difference in the worth for Control Bank B in cycle 6 is also the result of a slight mis-prediction in the local power distribution and possibly measurement error. However, the observed differences are well within the test acceptance criteria for individual bank worths of +/-30% or 200 pcm, whichever is greater.

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Attachment 2 (Non-Proprietary)

Response 2c:

A 10CFR 50.59 evaluation was performed to determine if any Unreviewed Safety Questions (USQs) exists when the current methodo!ogy is applied to a fuel design that differs from those previously benchmarked and documented in topical report DPC-NE-1004A. For the Duke Power Westinghouse designed nuclear plants, DPC-NE-1004A is considered applicable to Westinghouse OFA, Standard, and FCF Mark-BW (similar to Westinghouse Standard) fuel. The November i992 SER to this topical stipulated that "the application of CASMO-3 and SIhfULATE-3P tofuel designs that difer significantlyfrom those included in the topical data base should be supported by additional code validation to ensure that the DPC NE-1064 methodology and uncertainties apply." The fuel type evaluated in this 10CFR 50.59 evaluation was the Westinghouse Page 5

Attachment 2 (Non Proprietary)

Performance Plus fuel type (similar to Westinghouse Standard and RFA fuel) with integral fuel burnable absorber (IFBA). The integral fuel burnable absorber consists of a thin coating of ZrB 2 applied directly to fuel pellets of selected fuel rods. The analysis is applicable to the Westinghouse RFA fuel design as discussed in the answer to question 2b.

The results of the evaluation concluded that the methodology described in DPC-NE-1004A is applicable to fuel containing IFB A coated fuel pins. This conclusion is based on the results of benchmark calculations that showed code performance commensurate with that described in DPC-NE-1004A. Power distribution uncertainty factors calculated for fuel containing IFB A coated fuel rods, based on a 95% probability and confidence level, were bounded by uncertainty factors approved by the NRC in DPC-NE-1004A. Consequently, the introduction of IFBA fuel will not change the power peaking uncertainties assurned in the analysis of Updated Final Safety Analysis Report (UFSAR) Chapter 15 accidents. Therefore, it can be concluded from a nuclear design perspective that the consequences of UFSAR accidents previously evaluated are not increased and the margin to safety as defined in the bases to Technical Specifications is not decreased. In addition, safety margin will be maintained in future analyses through the application of a conservative combination of uncertainty factors. There are no USQs associated with this change.

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4 d Attachment 2 (Non Proprietary) l

3. Section 3.2 states that (1)in all nuclear design analysis,Imth the RFA and the Mark BW fuel are explicitly modeled in the transition cores, and (2) when establishing Operating and reactor protection system limits (i.e., loss of coolant accident (LOCA) kw/ft, l departure from nucleate boiling (DNB), centerline fuel melt (CFM), transient strain), l the fuel specific limits or a conservative overlay of the limits are used. Please elaborate I on the mixed core model for nuclear design analyses, and how fuel specific limits are used.

Response

The mixed core model used in the evaluation of transition cores containing RFA and Mark-BW fuel is based on the same methodology that is used to setup a nuclear model for a reactor cores containing a single fuel type. A SIMULATE-3 model is developed for each reload core design in accordance with the methodology described in DPC-NE-1004A. For mixed cores, this model contains cross sections and few group constants for each unique combination of fuel type (ie.

RFA or Mark-BW), enrichment and burnable poison loading and geometry. Cross sections and few group constants are derived from [ ] CASMO-3 calculations. The SIMULATE-3 model is used to confirm the acceptability of key physics parameters assumed in .

UFSAR Chapter 15 accident analyses and to develop core power distributions used in the I evaluation of LOCA, DNB, transient strain and centerline fuel melt limits.

l The generation of core power distributions for the development of core operational axial flux difference (AFD) limits and the f(AI) portion of the over-power delta-T and over-temperature delta-T trip functions (i.e. RPS limits) are conservatively performed using SIMULATE-3 based on the methodology described in DPC-NE-201 IPA. The power distributions developed during this process are compared against fuel specific Mark-BW and RFA LOCA, DNB, CFM and transient strain limits by assigning specific Mark-BW and RFA limits to each fuel type. Mark-BW and RFA fuel limits are developed using NRC approved methodologies. If positive margin exists to all limits, then no changes are made to operational AFD, or the RPS limits used in the development of the f(AI) trip functions. If any of the limits are exceeded, then either (1) the AFD or RPS limits are reduced to produce positive margin to all limits, (2) a specific analysis is performed on the out-of-limit parameter, or (3) the core is redesigned.

i In some instances it may be desirable to develop a single composite set of limits that can be used l to evaluate both fuel types. For this scenario, a conservative overlay of Mark-BW and RFA limits would be performed to develop a single set of limits that would be applicable to both Mark-BW and RFA fuel. Either of the above mentioned approaches is equally valid.

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, a Attachment 2 (Non Proprietary)

4. Section 5.2 states that in using thre 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. Willit he used for future RFA reload analyses?

(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?

Response Ja; The reference power distribution given in DPC-NE-2004P-A, Rev. I will be used in all future RFA analyses. This radial pin power distribution (the relationship of the peak pin to the remaining fuel pins in the highest power fuel assembly) used in DPC-NE-2009P and previous topical reports will not be modified. This maintains the relative radial power distribution the same as previously approved. There are no plans to change this distribution.

The peak pin value, however, could be increased in the future to utilize the increased thermal performance available in the RFA design. For DNB analyses using the Maximum Allowable Peaking (MAP) methodology described in DPC-NE-2004P-A, Rev.1, the key DNB parameter is the reference power distribution, not the peak pin power. The peak pin power is only meaningful when all other DNB parameters are specified (axial peak location and magnitude, core power level, RCS pressure, flow rate, and temperature). The reference power distribution is used to create the Maximum Allowable Peaking (MAP) limits that ensure the required level of DNBR protection is provided. The MAP limits define the maximum allowable peak pin as a function of axial peak. The reference power distribution is used consistently in all DNB analyses (core DNB limit lines, transient analyses, SCD statepoint determinations, etc.). Any change in the peak pin value will be evaluated in all DNB analyses and will be reflected in the Maximum Allowable Peaking limits provided in the COLR for each reload cycle.

The ability to increase the peak pin value is a result of a new fuel design, additional design features, a new or modified CHF correlation, or changes to the analysis conditions. If the performance improvement is related to fuel hardware or correlation change, a submittal is made to the NRC and approval required prior to use, if the change is to the analysis conditions and no methodology is modified, the change can be implemented through the 10CFR50.59 process. In either case, any increase in the peak pin value is not made unless all analyses and related licensing limits are verified to be conservatively satisfied.

Response 4b:

The reference power distribution used to create the Maximum Allowable Peaking (MAP) limits is used in all steady state generic analyses. This distribution is verified each reload by performing DNB calculations with cycle speGic predicted radial pin power distributions. This specific pin distribution comparison between what is predicted for a particular cycle and the generic analysis reference power distribution verifies the conservatism of the reference distribution.

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l Attichment 2 (Non Proprietary)

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 )

i 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 CIIF 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 HW test results as the DNHRs of the tests calculated with this combination yleided 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 wi!! 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 HW and RFA fuel designs.

l Response Sa:

The EPRI correlations will only be used in the RFA models in VIPRE-01. The Levy / Zuber-Findlay combination will be used when modeling Mark-BW fuel.

J l Duke considers the selection of the two-phase flow correlations to be a very minor effect on l DNBR analyses. Mark-BW CHF test data was analyzed with both Levy /Zuber-Findlay and EPRl/EPRI in the same manner as the RFA with comparable results.

Response Sh:

See 5(a) above.

Response Sc:

The transition core models use the simplified (8 Channel) models to maximize the impact of different fuel types. In the transition core model, the limiting assembly is modeled as an RFA and the rest of the core is modeled as Mark-BW fuel. Since the MDNBR occurs in the limiting assembly, the void correlations are input for the fuel type modeled as the limiting assembly. For the RFA/ Mark-BW transition core model, the RFA design is the limiting assembly; thus the EPRI set of correlations are used.

The transition analyses covered a wide range of statepoint fluid conditions and 3-dimensional core power distributions. This matrix of conditions were analyzed using both the EPRI and i Levy /Zuber-Findlay correlations with minimal difference in transition core results using either set of void correlations (average difference of <l% in peaking).

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Attachment 2 (Non-Proprietary)

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 HW fuel and the RFA design, and determine a conservative DNHR penalty to be applied for the transition cores. Table 5-4 presented the statistical DNHRs 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 HW transition cores for the 5000 case runs.

l (a) Why is the statistical design limit value pr@rietary information?

(h) With respect to the statistical core design methodology, describe how the uncertainties of the CilF correlation and the VIPRE code /model are propagated with the uncertainties of the selected parameters of each stalepoint for the calculafsn of the statistical DNHR 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 DNHR penalty factor?

Response 6a:

The Statistical Design Limit (SDL) will be changed to non-proprietary. This change will be included when the approved versions of the report are issued.

Response 6b:

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When a statepoint is selected, all key parameters, including CliF correlation and code /model uncertainties, are randomly varied based on the uncenainty distribution and magnitude. The resulting values of power, pressure, temperature, llow, and 3-D power distribution are used to create the VIPRE-01 input for the cases. After the code is executed and the DNBR calculated for l each case, the DNBR value is multiplied by the propagated values for the CHF correlation  !

l uncenainty and the VIPRE code /model uncertainty. This final DNBR value for each case (500 or 5000 cases are run for each statepoint) is used to determine the statepoint's statistical DNBR value.

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Response 6c:

l The analysis discussed in the last paragraph of Section 5.7 verified that the statistical DNB limit developed with a full core RFA model is valid for transition RFA/ Mark-BW cores. The limiting j statepoint (12TR) was evaluated using the RFA/ Mark-BW transition core model, confirming that the same statistical design limit can be used for transition and full core analyses.

l The transition core DNB penalty factor is determined separately using the RFA/ Mark-BW transition core model described in Section 5.7. The DNB penalty is determined by evaluating the effect of the transition core hydraulic behavior on the Maximum Allowable Peaking (MAP) limits l calculated for a full RFA core. The resulting DNB penalty is then accounted for in all l RFA/ Mark-BW transition core DNB analyses.

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Attichment 2 (Non Proprietary)

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 fucI 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 wcre performed with applicable form loss coefficients according to the vendor. Table 5.1 provides general RFA fuel specifications and -

characteristics without the hydraulic characteristics of the spacer grids.

(a) Provide comparisons for the thickness, height, and form loss coefficients of the RFA and Mark HW 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 CilF test assemblies if they are different from the values described in item (a).

(c) Describe the procedures to ensure that the form loss coefficients of the RFA grids are comparable to those used in the statistical core design analysis and the CilF tests so that both the WRH 2M CilF correlation DNHR limit and the statistical core design limit are valid.

Response 7a:

The grid data is shown in the following table:

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e O Attachment 2 (Non-Proprietary)

Response 7b:

The RFA CliF tests used Mixing Vane (MV) and intermediate flow mixing (IFM) grids representative of the production RFA design fuel assembly. The CIIF test sections are a 5x5 rod bundle with either all typical (unit) cells or typical cells with a thimble (guide tube) cell in the center. The form loss coefficients for the CliF test section are calculated for these subchannels and are based on the total 5x5 bundle flow area. Likewise, the fuel assembly subchannel form loss coefficients are calculated based on the fuel assembly flow area. The ratio of thimble / typical cell form loss coefficients, to which DNilR is sensitive, is equivalent for the CliF test section and the production grid (for both MV and IFM grids). Therefore, the CliF test section and production RFA grids are identical with respect to DNBR analyses.

In comparing the test versus production geometry, the vanes and strap features of the respective grid types are consistent. There is one slight difference between one of the CHF test sections and the production fuel assemblies. The thimble OD was 0.474 inches for the thimble CliF rod bundle section tested. The productior, assembly will have thimbles with an OD of 0.482 inches.

The difference in thimble tube OD has negligible impact on the correlation's predictive capability. This difference was addressed in WCAP-15025 and determined to be acceptable.

Response 7c:

The RFA analysis was completed with the form loss coefficients supplied in response to Question 7a. The transition core analysis used the RFA and Mark-BW values listed in the table in the respective model locations to accurately capture the hydraulic differences betw'een the fuel types side-by-side incore.

For each batch of fuel manufactured, critical RFA grid dimensions and form loss coefficients are supplied by the vendor to Duke Power. This data, along with other critical reload analysis parameters, are transmitted to Duke, on a batch basis, in a QA document known as the Databook.

Upon receipt of the Databook, the fuel design is frozen and may not be changed without Duke Power concurrence. This design notification process, including the process for changes occurring after the batch is frozen, is described in Duke Power Nuclear Engineering Workplace Procedure XSTP-101. The batch specific design information, transmitted in the Databook, will be used to ensure the validity of the Duke VIPRE-01 RFA models and associated SCD limit.

Any changes in the design data will be evaluated to verify that the generic analyses remain valid or the analyses will be revised using the new design data.

Page 12

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Attachment 2 (Non Proprietary)

8. Section 6.1.3 states that the thermal-hydraulic methodology described in DPC-NE 3000 PA Revision 1, with a simplified core model will be used for thermal-hydraulic analysis for the Updated Final Safety Analysis Report 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 anal 3ses 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 w hat additional conservatism is provided in using the simplified core model of DPC-NE-3000-PA.

(h) What is the criterion / criteria used to determine if transition core transient analyses are warranted? Ilow would it be determined that the criteria have been exceeded without RFA transition core analyses?

Response Sa:

The additional conservatism provided in using the simplified core model of DPC-NE-3000-PA is described in detail in Section 3.3.4 of DPC-NE-3000-PA (Reference 6-1 of DPC-NE-2009-P).

Response 8b:

Section 6.1.3 states the following. "No transition core transient analyses are performed as the results determined in Chapter 5 also apply for transient analyses. . . . Should it be determined in the future that transition core transient analyses are warranted, they will be performed accordingly." These statements summarize the results of an evaluation that has concluded that based on current information there is no need for performing transition core analyses for transients. The transition core effects on core thermal-hydraulic analyses for transients are adequately assessed by the steady-state core thermal-hydraulic transition core analysis in Chapter

5. The purpose of the second sentence quoted above was to state Duke's intent to evaluate any emerging issues or information, and, if necessary, to re-evaluate the current conclusion that no transient analyses of transition core effects are necessary. Duke does not expect any emerging information to change this conclusion, but Duke will address any such situations in the future.

Page 13

. c Attachment 2 (Non Proprietary)

9. Regarding rod ejection analysis using SIMULATE-3K, Section 6.6.2.2.1 states that the transient response is made more conservative by increasing the fission cross sections in the ejected rod h> cation 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 delayed neutron fraction, and i ejected rod worth, etc.

l l

(a) What are the values of the multiplication factors used for fission cross sections, and how are they determined?

(b) llow 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 he moderator temperature coefficients?

Response 9a:

An iterative process is used to determine the [ i

) Note that these multipliers are [

]

The methodology used to determine the [ ] adjustments is consistent with the power distribution adjustment methodology described in DPC-NE-3001 with one exception. [

1 The BOC and EOC [ ] multipliers are shown in Figures 9-1 and 9-2.

Figure 9-1 IlOC[ ] Multipliers H G F E D C B A 8 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 9 1.0 1.0 1.0 1.0 10 1.0 1.0 1.0 10 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 11 1.0 1.0 1.0 1.1450 0.9810 1.0445 1.0 1.0 12 1.0 1.0 1.0 0.9810 1.0904 0.9435 1.0 13 1.0 1.0 1.0 1.0445 0.9435 1.1358 1.0 14 1.0 1.0 1.0 1.0 1.0 1.0 15 1.0 1.0 1.0 1.0 Page 14

, . I Attachment 2 (Non Proprietary)

Figure 9 2 EOC[ ] Multipliers 11 G F E D C B A i 8 1.0 1.0 1.0 1.0 1.0 1.0 1.0

' 1.0 9 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 10 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 11 1.0 1.0 1.0 1.1955 0.9760 1.0960 1.0 12 1.0 1.0 1.0 _

1.0 0.9760 1.1230 0.9530 1.0 13 1.0 1.0 1.0 1.0960 0.9530 1.1240 1.0 l 14 1.0 1.0 1.0 1.0 1.0 1.0 15 1.0 1.0 1.0 1.0 Response &

The input multiplier " VAL" in equation 6.1 is determined through an iterative process until bounding control rod worths (ejected and trip rod worths), Doppler temperature coefficients and moderator temperature coefficients are determined. Unique multipliers are required for each of the parameters adjusted. As a result, unique sets of multipliers are calculated for each of the four rod ejection accident cases evaluated (ie. BOC liFP and HZP and EOC liFP and 11ZP).

Conservative Doppler temperature coefficients are calculated [

] For this case, the "X" variable in equation 6.1 is fuel temperature. The parameter i " VAL" is adjusted until a conservative Doppler temperature coefficient is determined.

I Conservative moderator temperature coefficients are developed by [

] Iterations are performed until a multiplier is determined that yields the desired moderator temperature coefficient. For this calculation, the X variable in equation 6.1 is moderator temperature.

A similar process is used to develop limiting control rod worths. Ejected rod wonhs are conservatively calculated by [

] Trip rod worths are minimized to conservatively limit the amount of negative reactivity inserted into the core post trip assuming the highest wonh control l rod and ejected control rod are fully withdrawn. For these cases, the X variable in equation 6.1 is l the [ ] Iterations are performed until a conservative ejected rod worth and trip rod worth are calculated.

The multiplier required to produce a conservative beta-effective is determined by re-arranging equation 6.2 to the following.

i l

Page 15 l

Attachment 2 (Non Proprietary)

The multipliers calculated for each of the key physics parameters assumed in each of the four rod ejection accidents (ie. BOC liFP and IIZP and EOC liFP and ilZP) are shown below. These multipliers were developed to produce bounding key physics parameters to ensure a conservative transient response. The multipliers presented are unique to each of the accidents presented in this report [

]

BOC liFP BOC llFP BOC llZP BOC llZP Parameter Multiplier (VAL) Target Value 51ultiplier (VAL) Target Value DTC 0.689 -0.90 pcm/ F 0.555 -0.90 pcm/ F MTC -0.005 0.0 pcm/ F -1.247 0.0 pcm/ F Ejected Rod Worth 1.168 200 pcm 1.029 720 pcm Trip Wonh 0.510 250 pcm I.650 250 pcm Beta-effective 0.882 0.0050 0.878 0.0050 ,

EOC liFP EOC liFP EOC llZP EOC llZP Parameter Multiplier (VAL) Target Value Multiplier (VAL) Target Value DTC 0.810 -1.20 pcnd F 0.666 -1.20 pcm/ F MTC 0.283 -10.0 pcm/ F 0.478 -10.0 pcm/ F Ejected Rod Wonh 1.055 200 pcm 0.868 900 pcm Trip Worth 1.073 250 pcm 1.650 250 pcm l Beta-cffective 0.768 0.0040 0.763 0.0040 l Response 9c:

No. The X's in equation 6.1 are moderator temperature. Refer to answer "9b" for additional information.

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Attichment 2 (Non Proprietary) l

10. Regarding the SIMULATE-3K code, there is an optional " frequency transform" approach, under the " Temporal Integration Models," that can be chosen to 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?

(h) What are the consequences of exercising (or not exercising) this option? Please provide technical, justification and comparisons of results.

Response 10a:

The frequency transform method is SIMULATE-3K's default transient neutronics solution option and was used in all of the transient evaluations presented in DPC-NE-2009. This approach was used because it is computationally more efficient and reproduces the results of finite difference methods, which require smaller times step to achieve the same accuracy as the frequency transform method.

The method used to solve the transient neutronics equations is determined by the code user and used throughout the transient. There is no switching of solution methods during the transient, l

Response 10b:

There are no physical consequences from using either the frequency transform or finite difference methods to solve the transient neutronic equations since both methods are equally accurate. From theory, the flux variation from one time step to the next is exponential. The frequency transform method takes credit for this behavior, instead of an assumed linear variation in simple finite difference methods. By taking credit for the exponential flux variation, computational efficicncy is increased because larger time steps can be taken without loss in accuracy as is the case in traditional methode. Therefore, the frequency transform method is the preferred solution technique because it produces the same answers as finite difference methods, but with reduced code execution time.

Sensitivity studies performed showed no difference in the peak core power or the time of the peak core power for cases where the frequency transform method was turned on and off. Therefore, it can be concluded that there is no consequence of using the frequency transform approach.

l

(

Page 17

j 1

Attochment 2 l

(Non Proprietary)

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 i reload design, physics, thermal-hydraulics, and transients and accidents, which wcre  !

approved by NRC for analyses of current Catan ha cores not having the RFA design.

For example, DPC-NE-1904A, DPC NE-2011 PA, DPC-NF-2010A and DPE-NE-3001-l 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 codc are used for non LOCA transient and accident analyses. Westinghouse small- and large break LOCA evaluation model described in WCAP-10054-P A and WCAP-10266 P A, and related topical reports, are used for the small and large-break LOCA analyses. Some of these methodologies have inherent limitations, and tome have l conditions or limitations imposed by the NRC safety evaluation reports in their ,

applications. Provide a list of the inherent limitations, conditions, or restrictions l l applicable to the RFA core design from all the methodologies to be used for the RFA l

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.

Response

DPC-NE 1004A, Duke Power Company Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P, Rev.1, April 1998.

The SER states that this methodology is acceptable for performing reload analyses for B&W 177 and Westinghouse 193 assembly reactor cores, subject to the following restrictions:

a. The application of CASMO-3 and SIMULATE-3 to fuel designs that differ significantly from those included in the topical data base should be supported by additional code validation to ensure that the DPC-NE-1004A methodology and uncertainties apply.

Resohstion: While Duke does not consider the introduction of the Integral Fuel Burnable Absorber (IFBA) in the Westinghouse RFA design to be a significant fuel design change, a conservative approach was adopted to confirm the acceptability of current nuclear uncertainty factors because of the availability ofIFBA benchmark data. The uncertainty analysis (described in Section 3.2 and in the answer to question 2) confirmed the acceptability of the currently licensed nuclear uncertainty factors for FAH, Fq and FZ for Westinghouse fuel containing IFBA and WABA burnable absorbers.

b. The system of codes represented in the topical report must be protected with appropriate quality assurance procedures, subject to auditing by the NRC staff.

Resoh< tion: The codes represented in the topical report DPC-NE-1004A are procedurally controlled and are in compliance with the Duke Energy Corporation Quality Assurance Topical Report which is in compliance with the requirements of 10CFR 50, Appendix B and other approved industry standards such as ANSI N45.2-1971 and ANSI N18.7-1976.

l Page 18 1

Attachment 2 (Non Proprietary)

DPC NE-201 IPA," Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990.

The SER for this methodology imposes the following restrictions:

a. The application of this methodology is limited to the McGuire and Catawba Nuclear Stations.

l Resolution: Duke is only using this methodology for McGuire and Catawba

b. The application of this methodology to other Westinghouse plants would be acceptable provided that plant-specific differences be considered and justified.

Resolution: Duke is only using this methodology for McGuire and Catawba. The use of this methodology for application to another Westinghouse unit (or units) would require the submittal of license amendments and NRC approval.

c. Application of this methodology is contingent upon NRC approval of the Reload Design Thermal-Hydraulic Methodology DPC-NE-2004P-A using the VIPRE-01 code. (Topical Approved) l Resolution: The Thermal-Hydraulic Methodology described in Topical Report DPC-NE-2004P-A has been approved.

d. Calculation of power and xenon distributions are limited to the use of the EPRI-NODE-P and the PDQ-07 codes Resolution: The approval of the Topical Report DPC-NE-1004A allowed the substitution of l either the CASMO-3/ SIMULATE-3P or the CASMO-3/ NODE-P codes in place of the EPRI-NODE-P and PDQ-07.

l DPC-NF 2010A," Duke Power Company McGuire Nuclear Station Cataw ba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985.

The SER for this methodology imposes the following restrictions:

None, with the exception that the methodology in sections 6.3,7.1,7.2,7.3,7.4 and 7.4.1 were excluded from this report. The replacement methodology is described below:

a. Section 6.3: Comparison of Cycle Specific Safety Related Physics Parameters l Resolution: This methodology was replaced by the methodology described in Topical Report l DPC-NE-3001PA.

l l b. Section 7.1 - 7.4.1: Three-dimensional peaking analysis Resolution: This methodology was replaced by the methodology described in Topical Report DPC-NE-2011PA.

Page 19

Attachment 2 (Non-Proprietary)

DPC NE 2004P-A," Duke Power Company McGuire and Catawha Nuclear Stations Core Thermal llydraulic Methodology using VIPRE 01," Revision 1, February 1997.

The limitations, conditions or restrictions identified in the SER and TER for DPC-NE-2004, Revision 1, are:

l

a. The DPC developed statistical core design methodology, as described in the submittal, is a generic methodology and is conceptually acceptable and generally applicable to other PWR plants; however, the approval we recommend at this time is only for McGuire and Catawba Nuclear Stations due to DPC's use of the specific uncertainties and distributions based upon plant data and its selection of statepoints used for generating the statistical design lirnit.

Resolution: The RFA SCD analysis presented in DPC-NE-2009 is only for McGuire and Catawba.

b. Either the response surface model (RSM) must be re-evaluated or the " simplified method" for determining an SDL using VIPRE-01 directly must be used whenever any of the following occur:

l a significant change is made in the fuel assembly design

! e a new or revised CilF correlation is developed e

operating conditions outside the range of conditions considered in the development of the RSM l The licensee is further required to make a submittal to the NRC for review if a new SDL is l calculated as a result of conditions outside the range of conditions considered in the 1

development of the RSM.

Resolution: The RSM was not used to calculate the SDL in DPC-NE-2009. The RFA analysis presented in DPC-NE-2009 calcuhtes the SDL for the RFA fuel with the WRB-2M j CliF correlation as per the " simplified method" referenced in DPC-NE-2004, Rev 1 which is the SCD calculation methodology subsequently approved in DPC-NE-2005, Rev 1.

c. Whenever DPC intends to use other CHF correlations, power distribution, fuel pin conduction model, or any other input parameters and default options which were not part of l the original review of the VIPRE-01 code, DPC must submit itsjustificat; for NRC review and approval.

Resolution: DPC-NE-2009 identifies the VIPRE-01 modeling requirements as well as the CliF correlation and statistical analysis limit.

d. Core bypass flow is cycle dependent. DPC will verify, in future applications, that its use of a particular core flowrate resulting from a bypass flowrate for that cycle is bounded by the range of values used in the subject topical report. Otherwise DPC will reassess the need for regeneration of a new response surface model.

Page 20

i 1

Attachment 2 (Non Proprietary)

Resolution The value of core bypass now used in the generic RFA SCI) analysis is expected to bound the values for all reload cores. The core bypass flow will be verified on a cycle by cycle basis to ensure conservatism.

1 DPC-NE 2005P-A, Thermal-flydraulic Statistical Core Design Methodology", Revision 1, November 1996.

l The limitations, conditions or restrictions identified in the SER and TER for DPC-NE-2005P-A are:

a. The statistical core design (SCD) methodology developed by DPC, as described in the .

submittal (DPC-NE-2005), is direct and general enough to be widely applicable to any )

pressurized-water reactor (_PWR) fuel or reactor, provided that the VIPRE-01 methodology is approved with the use of the core model and correlations including the critical heat flux (CliF) correlation subject to the conditions in the VIPRE safety evaluation report (SER).

DPC committed in their topical report that its use of specific uncertainties and distributions will be justified on a plant specific basis, and also that its selection of statepoints used for generating the statistical design limit will be justified to be appropriate. The methodology is approved only for use in DPC plants.

Resolution: Addressed in Chapter 5 of DPC-NE-2009. The RFA analysis presented in DPC-NE-2009 is only for McGuire and Catawba.

h. Of the two DNBR limits, only the use of the single, most-conservative DNBR limit is approved.

Resolution: Use of two DNBR limits was not requested in this submittal. The single DNBR limit stated for use for RFA fuel in full cores or transition cores will be used for all statepoints within the conditions listed in Table 5-5.

WCAP-15025," Modified WRB 2 Correlation, WRB 2M, for Predicting Critical IIcat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids".

The limitations, conditions or restrictions identified in the SER and for WCAP-15025 are:

a. Since WRB-2M was developed from test assemblies designed to simulate Modified 17x17 Vantage 511 fuel the correlation may only be used to perform evaluations for fuel of that type without furtherjustification. Modified Vantage 5H fuel with or without modified intermediate flow mixer grids may be evaluated with WRB-2M.

Resolution: The SCD analysis presented in DPC-NE-2009 is for the RFA design, which includes the modified low pressure drop structural mid-grids and modified intermediate flow l

mixing grids (see Chapter 2 of DPC-NE-2009). The WRB-2M CIIF correlation is used for all DNBR calculations on the RFA.

l Page 21 l

Attachment 2 (Non Proprietary) l

b. Since WRB-2M is dependent on calculated local fluid properties these should be calculated by a computer code that has been reviewed and approved by the NRC staff for that purpose.

Currently WRB-2M with a DNBR limit of 1.14 may be used with the TlHNC-IV computer code. The use of VIPRE-01 by Westinghouse with WRB-2M is currently under separate review.

Resolution: The DNB analyses in DPC-NE-2009 are performed with VIPRE-01. As stated in Section 2.3 and 3.3 of WCAP-15025, bott VIPRE-Ol and THINC-IV were used to analyze the CliF test data. Tables A-1 to A-4 in the /.ppendix to WCAP-15025 show the local fluid conditions calculated with VIPRE-01. Also, the SER for WCAP-15025 states that the "results of the TlilNC-IV analyses agreed with those from VIPRE-01." Additionally, VIPRE-01 was approved for use in thermal / hydraulic analyses at McGuire and Catawba in 1

DPC-NE-2004, Revision 1. Based on this, Duke has used and will continue to use VIPRE-01 to perform all RFA analyses.

c. WRB-2M may be used for PWR plant analyses of steady state and reactor transients other than loss of coolant accideats. Use of WRB-2M for loss of coolant accident analysis will require additional justification tint the applicable NRC regulations are met and the computer i code used to calculate local fuel element thermal / hydraulic properties has been approved for that purpose.

Resolution: The RFA LOCA analysis is not described in DPC-NE-2009. The LOCA analysis is performed by the fuel vendor with the approved correlations specified by the vendor's methodology. The CliF correlation used in the LOCA analysis is listed in WCAP-8301.

d. The correlation should not be used outside the range of applicability defined by the range of the test data from which it was developed. This range is listed in Table 1.

Resolution: Table 1 is listed below for reference.

l Parameter Range Pressure (psia) 1495 s P s2425 Local Mass Velocity (Mibm/hr-ft*) 0.97s G $3.1 Local Quality -0.1 s X s 0.29 lieated length, inlet to CliF location (ft) Lhs14 Grid Spacing (in) 10 s gsp s 20.6 Equivalent hydraulic diameter (in) 0.37 s De s 0.46 i Equivalent heated diameter (in) 0.46 s Dh s 0.54 1

The WRB-2M C11F correlation was used for all RFA DNBR calculations. The fluid parameter ranges (first two items on the parameter list) are confirmed by the statepoint selection listed in Chapter 5 of DPC-NE-2009. The fuel design related parameters (last four items on the parameter list) are confirmed implicitly by the fuel model. The local quality limit is verified for each analysis.

l Page 22

Attachment 2 (Non Proprietary)

DPC NE-3000 PA," Thermal Ilydraulic Transient Analysis Methodology," Revision 2, December 1997.

The original SER dated November 15,1991 lists the following conditions (Section 3.0). The SER for Revision 1 dated August 8,1994 does not have any limitations or conditions for McGuire and Catawba. The SER for Revision 2 dated October 14,1998 does not have any new limitations or conditions.

a.

With respect to analyzing transients which result in a reduction in steam generator secondary water inventory, use of the RETRAN-02 steam generator modeling is acceptable, only for transients in that category for which the secondary side inventory for the effective steam generator (s) relied upon for heat removal never decreases below an amount which would cover enough tube height to remove decay heat.

Resolution: By letter dated September 25,1998 (G. R. Peterson to NRC Document Control Desk), Duke notified the NRC of a new RETRAN-02 steam generator model which addresses this SER condition for the Catawba Unit 2 UFSAR Section 15.2.7 loss of normal feedwater analysis. This submittalis currently under NRC review. The subject of this condition is not applicable for all other UFSAR Chapter 15 transients and accidents for McGuire and Cataw ba.

b. All generic limitations specified in the RETRAN-02 SER.

Resolution: By letter dated June 3,1991 (M.S. Tuckman to NRC Document Control Desk),

Duke responded to the generic limitations specified in the RETRAN-02 SER in the response to Question #29. This response along with subsequent methodology revisions (including the revisions in DPC-NE-2009-P) have all been submitted to the NRC. Later RETRAN-02 SERs were reviewed and it was determined that three new SER conditions exist for the RETRAN-02 MOD 005.0 code version (SER dated November 1,1991). The responses to these conditions are as follows:

i

1) The user must justify, for each transient in which the general transport model, the selected degree of mixing with considerations as discussed in Section 2.1 of this SER.

l l Response: Topical report DPC-NE-3001-P described the application of the general j transpon model in the Duke methodology. The topical report was reviewed and approved by the NRC.

2) The user must justify, for each use of the ANS 1979 standard decay heat model, the associated parameter inputs, as discussed in Section 2.2 of this SER.

i Response: The Duke modeling of decay heat as described in topical report DPC-NE- ,

3002-A is based on the ANSl/ANS-5.1-1979 standard plus a two-sigma uncertainty.

This decay heat modeling approach is standard in the industry for non-LOCA analyses, and meets the intent of this condition. The NRC has reviewed and approved DPC-NE-3002-A.

3) Because of the inexactness of the new reactivity edit feature, use of values in the edit either directly or as constituent factors in calculations of parameters for comparison to formal performance criteria must be justified.

Page 23

1 Attachment 2 I (Non Proprietary)

)

Response: The Duke methodology does not use the reactivity edit feature in the manner that is the subject of this condition. Therefore this condition is not applicable.

c. Determination of acceptability is based upon review of selection of models/ correlations for transients involving symmetric core neutronic and thermal-hydraulic conditions only. Thus.

VIPRE-01 models are approved for use in analyzing symmetric transients only.

Resolution: The DPC-NE-3001-PA topical report submitted VIPRE-01 models for transients involving asymmetric core neutronic and thermal-hydraulic conditions. NRC approval of DPC-NE-3001-PA closed out this condition.

d. When using the DPC developed SCD method, the licensee must satisfy the conditions set forth in the staff's safety evaluation of DPC-NE-2004.

1 Resolution: Duke responded to this question by letter dated August 29,1991 (M. S.

Tuckman to NRC Document Control Desk). Attachment 2 to this letter addresses the applicable conditions and how these conditions are met, and is summarized as follows. The first condition requiring submittal of models for asymmetric transients was met with the submittal of DPC-NE-3001. The second condition required a transition core penalty to be applied. The details of the transition core penalty modeling were presented in the response.

The third condition required modeling to avoid errors related to the use of the subcooled boiling models. A commitment to properly apply this model was made. The fourth condition required submittal of the BWCMV correlation prior to use, which was done. The fifth condition required future submittal of any methodology changes to important inputs and models such as different CHF correlations, power distributions, input options, etc. Duke observes this condition and has and will submit such methodology changes prior to implementation. The sixth condition requires that the core bypass flow be determined and justified on a cycle-by-cycle basis. Duke commits to confirming that the core bypass flow for each reload cycle will be bounded by the core bypass flow assumed in the analyses.

e. Wheneser DPC intends to use other CHF correlations, power distribution, fuel pin conduction model or any other input parameters and default options which were not part of the original review of the VIPRE-01 code, DPC must submit its justification for NRC review and approval.

1 Resolution: Duke recognizes the requirements of this condition and continues to meet this -

condition. For example, Revision 2 to DPC-NE-3000-P (Letter, M. S. Tuckman to NRC l

Document Control Desk, December 23,1997) submitted revised VIPRE-01 methodology to I include the Mk-B 11 fuel assembly design and the BWU-Z CHF correlation, along with other minor changes. Future revisio'ns to Duke topical reports will be submitted as necessary per j the requirements of this condition.

l Page 24

Attachment 2 l (Non Proprietary)

)

i l

DPC-NE 300lP-A," Multidimensional Reactor Transients and Safety Analysis Physics I Parameters Methodology," Novemher 1991. l The SER dated November 15,1991 lists the following limitations (Section 4.0).

a. The licensing application of the SIMULATE-3P static methods for determining the key safety parameters requires NRC approval of the reference topical repon, DPC-NE-1004 (Section 3.1)

I Resolution: Topical report DPC-NE-1004-A was reviewed and approved by the NRC. The latest NRC SER for DPC-NE-1004-A, Revision 1,is dated April 26,1996

b. The licensing application of the DPC-NE-3001-P transient analysis methods requires NRC approval of MOD 005 of RETRAN-02 for boron transpon calculations (Section 3.5)

Resolution: The NRC SER for RFTRAN-02 MOD 005.0 was dated November 1,1991.

c. The licensing application of the DPC-NE-3001-P transient analysis methods requires NRC approval of the thermal-hydraulics topical repon DPC-NE-3000 (Section 3.5)

Resolution: The NRC SER (McGuire/ Catawba scope) for DPC-NE-3000-PA was dated November 15,1991.

DPC-NE-3002-A,"UFSAR Chapter 15 System Transient Analysis Methodology," Revision 2, December 1997.

The original SER dated November 15,1991 lists the following conditions (Section 3.0). i 1

a. DPC's Statistical Core Design methodology treats seven state variables as key parameters. )

Four of these variables were accounted for in this topical report. Of the remaining parameters, the power factors are also input items for systems analysis, which was not presented in the topical report. Similarly, reactivity feedback was not discussed in this report.

Both of these parameters can significantly influence the course of the transient. Therefore, when application of the philosophical approach reported in this topical report is made and submitted for NRC review and approval, review should be made of the modeling of power and reactivity feedback, and to assure that such modeling has no adverse impact on the other modeling described herein.

Resolution: The power factors used in the models are described in the DPC-NE-3000 topical report, which has been reviewed and approved by the NRC. The reactivity feedback modeling is described in the DPC-NE-3001 topical report, which has been reviewed and

! approved by the NRC. The application of the integrated methodologies, including UFSAR Chapter 15 revisions, was submitted on June 26,1991 for the McGuire 1 Cycle 8 reload license amendment application. The SER for this submittal was dated November 27,1991.

The above DPC-NE-3002-A SER condition appears to be directed at the NRC review of the other topical reports and to the application of the methodology. It is inferred via NRC review and approval of all of the related topical reports and of the McGuire 1 Cycle 8 reload that the intent of this condition has been met.

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b. Validity of DPC's assumption of 120(7c of design pressure as part of the acceptance criteria for Reactor Coolant Pump Locked Rotor should be determined by the NRC staff.

l Resolution: Duke has adopted an acceptance criterion of 110'7c of design pressure for the locked rotor accident analysis as stated in Section 4.3 of DPC-NE-3002, Revision 2.

c. No justification was presented for trip and actuation times assumed in the Feedwater System I Pipe Break event analysis. Suchjustifications must be presented when this methodology is applied.

l Resolution: The NRC SER, Section 2.2.1, dated November 15,1991 specifically states that this TER condition is outside of the scope of DPC-NE-3002 and this review. Therefore, Duke has not prepared a response to this TER condition.

d. DPC documented intent to perform parametric studies in order to select conservative scenarios or assumptions throughout the subject topical report. Therefore, such parametric studies must be presented when this methodology is applied. l Resolution: The DPC-NE-3002-A topical report states that parametric studies are necessary to determine the conservative modeling approach for a limited number of assumptions for some of the transients. These parametric studies were performed and are documented in the engineering calculations. The results of the analyses using the conservative modeling approach and assumptions were submitted for NRC review with the McGuire 1 Cycle 8 reload license amendment request dated June 26,1991. These results were in the form of UFSAR revisions. Since it is not typical to include results of parametric studies in the UFSAR, only the results of the limiting cases were presented in the submittal of the application of the methodology. The engineering calculations which document the parametric studies are available for audit. It is concluded that this condition has been adequately addressed.

The SER for Revision I dated December 28,1995 has the following conditions in Section 4.0.

The SER for Revision 2 dated April 26,1996 does not have any new limitations or conditions.

a. The acceptability of the use of DPC's approach to FSAR analysis is subject to the conditions of SERs on all aspects of transient analysis and methodologies (DPC-NE-30(X), DPC-NE-3001, DPC-NE-3002, DPC-NE-2004, DPC-NE-2005) as well as the SERs on RETRAN and VIPRE computer codes.

Resolution: This condition has been addressed in this submittal.

b. There are scenarios in which an SGTR event may result in loss of subcooling and the consequent two-phase flow conditions in the primary system. In such instances, the use of )

RETRAN is not acceptable without a detailed review of the analysis. I I

Resolution: The McGuire and Catawba UFSAR SGTR analyses jo not result in a loss of subcooling.

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c. In the future if hardware or methodology changes, selection of limiting transients needs to be reconsidered, and DPC is required to perform sensitivity studies to identify the initial conditions in such a way to avoid conflict between transient objective, such as DNB and worst primary pressure.

Resolution: Duke's methodology, as described in DPC-NE-3002-A, does select initial and boundary conditions with consideration of the possibility that different selections and possibly separate analyses may be necessary depending on the acceptance criteria and the margin to the acceptance criteria. This approach will be continued for future re-analyses due to hardware or methodology changes.

d. It is emphasized that, when using the SCD methodology to determine DNBR, the range of applicability of the selected CHF correlation must not be violated.

Resolution: Duke recognizes the need to restrict the use of CHF correlations to within their ranges of applicability. Any deviations from this approach will be submitted for review and approval.

e. DPC's assumption of 1207c of design pressure as part of the acceptance criteria for Reactor Coolant Pump Locked Rotor is not acceptable. DPC is required to use i107c of design pressure for that limit.

Resolution: Duke has revised DPC-NE-3002-A to use i10% of design pressure as an acceptance criterion for the locked rotor accident (See Section 4.3 of DPC-NE-3002-A, Revision 2).

Westinghouse LOCA Topical Reports Westinghouse will provide the requested information regarding SER limitations, conditions, and restrictions for the LOCA-related topical reports referenced by DPC-NE-2009-P and to be used for the RFA and transitional RFA/ Mark-BW cores. This information will be submitted to the NRC by April 1,1999.

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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 CIIF correlation for the RFA design is 2425 psia, the 2400 psia safety limit line is within the range of the CIIF correlations for the Mark-BW and RFA fuel designs.

Ilowever, the safety limit lines in Figure 2.1.1 1 were based on the CIIF correlation for the Mark BW fuel design,in addition to the hot leg boiling limit. lias 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?

Response :

Yes. As stated, the 2400 psia line was selected since it was already defined for the Mark-BW fuel. Using the reference power distribution and the reactor inlet conditions defined by the hot leg boiling and DNB portions of the 2400 psia Safety Limit Line, the MDNBR was calculated using the full RFA core VIPRE-01 model and the WRB-2M CHF correlation. Additionally, the transition RFA/ Mark-BW cores were also evaluated to ensure the established limits were conservative. The MDNBR values were greater than the design DNBR limit for all of the cases in both evaluations.

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13. TS Surveillance Requirements (SRs) 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 Fai(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 Fai(x,y) has decreased since the previous surveillance. For the reactor core containing the RFA fuel design with integral burnable absorbers, a larger penalty may be required over certain burnup ranges early in the cycle due to the rate of 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 81 and 8-2, respectively, provide " typical values" for the burnup-dependent margin decrease penalty factors for the heat flux and enthalpy rise hot channel factors.

(a) Provide the actual values of the margin-decrease penalty factors, as well as the 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.

Response 13a:

l Margin decrease penalty factors will be calculated for each reload core. The actual margin l decrease penalty factors for the initial transition core can not be provided until the final design for l this core is complete. The cycle-specific factors for each core design will be included in each l units' cycle-specific Core Operating Limits Report (COLR).

l The methodology used to calculate the Fy (x,y,z) and Fai(x,y) margin decrease penalty factors is described below. The peaking factors used to calculate the margin decrease penalty factors are obtained from the analysis performed to establish operational axial flux difference limits as described in DPC-NE-201 IPA (" Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors).

Nuclear Heat Fha Hot Channel Factor, Fq:

Fy(x,y,z) is measured periodically using the incore detector system to ensure that the value of the total peaking factor, F y-RTP. assumed in the accident analysis is bounding. The frequency requirement for this measurement is 31 effective full power days (EFPD). In order to account for the possibility that Fy(x,y,z) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the measurement indicates that the Fy(x,y,z) measurement would exceed the Fy (x,y,z) limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval would be decreased based on the available margin, or the Fy(x,y,z) measurement would be increased by an appropriate penalty and compared against the Fy (x,y,z) operational and RPS surveillance limits to ensure allowable total peaking limits are not exceeded.

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The Fy(x,y,z) penalty factor is calculated by projecting the change in the [

]

p -

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] The Fq margin decrease factor may be applied directly to the measured Fq or may be incorporaint into the My(x,y,z) and Mc(x,y,z) margin factors as described in DPC-NE-201 IPA. For burnup ranges where the Fq margin decrease factor is less than 1.02, a value of 1.02 will be maintained.

Nuclear Enthalpy Rise Hot Channel Factor, Fag The nuclear enthalpy rise hot channel factor, Fm(x,y), is measured periodically using the incore detector system to ensure that fuel design criteria are not violated and accident analysis

assumptions are not violated. The frequency requirement for this measurement is 31 effective full power days (EFPD). In order to account for the possibility that Fm(x,y) may increase between surveillances, a trend of the measurement is performed to determine the point where peaking would exceed allowable limits if the current trend continues. If extrapolation of the i measurement indicates that the Fm(x,y) measurement would exceed the Fm(x,y) surveillance l

limit prior to 31 EFPD beyond the most recent measurement, then either the surveillance interval would be decreased based on the available margin, or the Fm(x,y) measurement would be increased by an appropriate penalty and compared against the Fm(x,y) surveillance limit to ensure allowable peaking limits are not exceeded. '

The Fm(x,y) penalty factor is calculated by projecting the change in the [

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] The Fm margin decrease factor may be applied directly to the measured Fm or may be incorporated into the Mm(x,y) margin factors. For burnup ranges where the Fa margin decrease factor is less than 1.02, a value of 1.02 will be maintained.

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Response 13b:

The methodology used to calculate the Fai(x,y) and F y(x,y,z) margin-decrease peaking penalty factors was described in answer 13a. Duke intends to reference this topical report (DPC-NE-2009) in Technical Specification 5.6.5 for the approved methodology used to calculate these l parameters.

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l Response to NRC Requests for Additional Information Dated December 9, 1998 and January 5, 1999 Applicable to s Duke Energy Corporation License Amendment Requests Dated July 22, 1998

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