ML16315A272
| ML16315A272 | |
| Person / Time | |
|---|---|
| Site: | Harris, Robinson  |
| Issue date: | 11/10/2016 |
| From: | Henderson K Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16315A286 | List:
|
| References | |
| DPC-NE-3008, RA-16-0034 | |
| Download: ML16315A272 (77) | |
Text
KELVIN HENDERSON Senior Vice President Nuclear Corporate 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-1295 Kelvin.Henderson@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED Serial: RA-16-0034 10 CFR 50.90 November 10, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23
SUBJECT:
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING APPLICATION TO REVISE TECHNICAL SPECIFICATIONS FOR METHODOLOGY REPORT DPC-NE-3008, REVISION 0
REFERENCES:
- 1. Duke Energy letter, Application to Revise Technical Specifications to Adopt Methodology Report DPC-NE-3008-P Revision 0, Thermal-Hydraulic Models for Transient Analysis, dated November 19, 2015 (ADAMS Accession No. ML15323A351)
- 2. Duke Energy letter, Supplemental Information for License Amendment Request Regarding Methodology Report DPC-NE-3008-P, dated October 3, 2016 (ADAMS Accession No. ML16278A080)
- 3. NRC letter, Requests for Additional Information Regarding Application to Adopt DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis (CAC Nos.
MF7112 and MF7113), dated October 24, 2016 (ADAMS Accession No. ML16216A061)
Ladies and Gentlemen:
In Reference 1, Duke Energy Progress, LLC (formerly referred to as Duke Energy Progress, Inc.), referred to henceforth as Duke Energy, submitted a request for an amendment to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and H. B.
Robinson Steam Electric Plant, Unit No. 2 (RNP). Specifically, Duke Energy requested NRC review and approval of DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis, and adoption of the methodology into the TS for HNP and RNP. In Reference 2, Duke Energy submitted a supplement to the amendment request that superseded Reference 1 in its entirety. In Reference 3, the NRC requested additional information (RAI) regarding this proposed amendment.
PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0034 Page 2 provides Duke Energy's response to the Reference 3 RAl's. Attachment 4 contains information that is proprietary to Duke Energy. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 4 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of Attachment 4. A non-proprietary version of Attachment 4 is included in Attachment 3. Attachment 5 is an Electric Power Research Institute (EPRI) document that is being provided to support the Attachment 4 response. Attachment 5 is proprietary to EPRI and is requested to be withheld from public disclosure in accordance with 10 CFR 2.390. An affidavit is included (Attachment 2) attesting to the proprietary nature of Attachment 5.
This submittal contains no new regulatory commitments. In accordance with 10 CFR 50. 91, Duke Energy is notifying the states of North Carolina and South Carolina by transmitting a copy of this letter to the designated state officials. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager-Nuclear Fleet Licensing, at 980-373-2062.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on November 10, 2016.
Sincerely, Kelvin Henderson Senior Vice President - Nuclear Corporate JBD Attachments: 1. Affidavit of Kelvin Henderson
- 2. Affidavit of Neil Wilmshurst
- 3. Response to NRC Request for Additional Information (Redacted)
- 4. Response to NRC Request for Additional Information (Proprietary)
- 5. EPRI RETRAN-30 Computer Code Manuals, NP-7450(A) (Proprietary)
PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED
PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-16-0034 Page 3 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 4 AND 5 THIS LETTER IS UNCONTROLLED cc:
(all with Attachments unless otherwise noted)
C. Haney, Regional Administrator USNRC Region II M. Riches, USNRC Senior Resident Inspector - HNP J. Zeiler, USNRC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP D. J. Galvin, NRR Project Manager - RNP W. L. Cox, III, Section Chief, NC DHSR (Without Attachments 4 and 5)
S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC)
(Without Attachments 4 and 5)
A. Wilson, Attorney General (SC) (Without Attachments 4 and 5)
A. Gantt, Chief, Bureau of Radiological Health (SC) (Without Attachments 4 and 5)
RA-16-0034 Affidavit of Kelvin Henderson RA-16-0034 Page 1 of 3 AFFIDAVIT of Kelvin Henderson
- 1. I am Senior Vice President of Nuclear Corporate, Duke Energy Corporation, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.
- 2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energys application for withholding which accompanies this affidavit.
- 3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in to Duke Energy RAI response letter RA-16-0034 regarding application to revise technical specifications for report DPC-NE-3008-P.
- 4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, 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 disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.
(ii)
The information is of a type that would customarily be held in confidence by Duke Energy. Information is held in confidence if it falls in one or more of the following categories.
(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by a vendor or consultant, without a license from Duke Energy, would constitute a competitive economic advantage to that vendor or consultant.
(b) The information requested to be withheld consist of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.),
and the application of the data secures a competitive economic advantage for example by requiring the vendor or consultant to perform test measurements, and process and analyze the measured test data.
(c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation assurance of quality or licensing of a similar product.
(d) The information requested to be withheld reveals cost or price information, production capacities, budget levels or commercial strategies of Duke Energy or its customers or suppliers.
RA-16-0034 Page 2 of 3 (e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.
(f) The information requested to be withheld consists of patentable ideas.
The information in this submittal is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for this declaration is the use of this information by Duke Energy provides a competitive advantage to Duke Energy over vendors and consultants, its public disclosure would diminish the informations marketability, and its use by a vendor or consultant would reduce their expenses to duplicate similar information. 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 Energy.
(iii)
The information was transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390, 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 is that which is marked in to Duke Energy RAI response letter RA-16-0034 regarding application to revise technical specifications for report DPC-NE-3008-P. This information enables Duke Energy to:
(a) Support license amendment requests for its Harris and Robinson reactors.
(b) Support reload design calculations for Harris and Robinson reactor cores.
(vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.
(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.
(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants 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 Energy.
- 5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.
RA-16-0034 Page 3 of 3 Kelvin Henderson affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on November 10, 2016.
~
RA-16-0034 Affidavit of Neil Wilmshurst
~~~1 1 ELECTRIC POWER t=l-1~
RESEARCH INSTITUTE Ref. EPRI Project Number 889-10 August25,2016 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Request for Withholding of the following Proprietary Information:
Neil Wilmshurst Vice President, Nuclear and Chief Nuclear Officer RETRAN-3D computer code manuals, NP-7450(A), Research Project 889-10, dated September 2014 To Whom It May Concern:
This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the documents identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI").
EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the submittal to the NRC by Duke Energy. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.
If you have any questions about the legal aspects of this request for withholding, please do not hesttate to contact me at 704-595-2732. Questions on the content of the documents should be directed to Kelli Voelsing of EPRI at 704-595-2878.
Sincerely, bu~
Attachment(s)
Together... Shaping the Future of Electricity 1300 West W.T. Harris Boulevard, Charlotte, NC 28262*8550 USA* 704.595.2732 *Mobile 704.490.2653
- nwilmshurst@epri.com
r!!!!!!r-::t~1 1 ELECTRIC POWER a=1-1c; RESEARCH INSTITUTE AFFIDAVIT of Neil Wilmshurst
- 1. I am Vice President, Nuclear and Chief Nuclear Officer of Electric Power Research Institute (EPRI), and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of EPRI.
- 2. I have knowledge of the criteria used by EPRI in designating information as proprietary or confidential. I am familiar with the EPRI information contained in the RETRAN-30 computer code manuals, NP-7450(A), Research Project 889-10, dated September 2014 and referred to herein as "Document". Information contained in this Document has been classified by EPRI as proprietary in accordance with policies established by EPRI for the control and protection of proprietary and confidential information
- 3. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."
- 4. The following criteria are customarily applied by EPRI to determine whether information should be classified as proprietary.
(i)
The information sought to be withheld from public disclosure is owned by EPRI and has been held in confidence by EPRI and its consultants.
(ii)
The information is of a type that would customarily be held in confidence by EPRI.
Information is held in confidence if it falls in one or more of the following categories.
(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by a vendor or consultant, without a license from EPRI, would constitute a competitive economic advantage to that vendor or consultant.
(b) The information requested to be withheld includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for EPRI.
(c) Use by a competitor of the information requested to be withheld would reduce the competitor's expenditure of resources, or improve its competitive position, in the design, manufacture. shipment, installation assurance of quality or licensing of a similar product.
(d) The information requested to be withheld is vital to a competitive advantage held by EPRI, would be helpful to competitors to EPRI, and would likely cause substantial harm to the competitive position of EPRI.
Together... Shaping the Future of Electricity 1300 Wesl W.T. Harris Boulevard, Charlolle, NC 28262-8550 USA* 704.595.2732 *Mobile 704.490.2653
- nwilmshursl@epri.com
The information in this presentation is held in confidence for the reasons set forth in paragraphs 4(ii)(a), 4(ii)(b) and 4(ii)(c) above. The information consists of analysis methodology details, analysis results, supporting data, and a method of analysis that provides a competitive advantage to EPRI.
(iii)
The information was transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390, it is to be received in confidence by the NRC.
(iv)
The information sought to be protected is not available in public.
- 5. Public disclosure of this information is likely to cause harm to EPRI because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing EPRI to recoup a portion of its expenditures or benefit from the sale of the information.
Neil Wilmshurst affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on August 25, 2016.
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Neil Wilmshurst (State of North Carolina)
(County of Mecklenburg)
Subscribed and sworn to (or affirmed) before me on this 25th day of August, 2016, by Neil Wilmshurst, proved to me on the basis of satisfactory evidence to be the person{s) who appeared before me.
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RA-16-0034 Response to NRC Request for Additional Information (Redacted)
Note: Text that is within double brackets (original NRC RAI wording) or brackets with an a,c superscript (Duke Energy response) is proprietary to Duke Energy and has been removed.
RA-16-0034 Page 1 of 65 NRC RAI 1 The NRC safety evaluation for DPC-NE-3000-PA stated that review of actual licensing applications and associated conservative assumptions was beyond the scope of the DPC-NE-3000-PA review, since such details are to be presented in a future topical report. It is not clear whether this is the intent in DPC-NE-3008-P, Section 4.3, which provides benchmark analyses and states that, "The analyses are not intended for direct incorporation into the HNP FSAR or RNP UFSAR [updated final safety analysis report] and are not being submitted for review and approval as new analyses of record (AORs)."
Clarify the licensing scope of the DPC-NE-3008-P methodology and clarify whether Duke Energy plans on submitting future analyses using this methodology for NRC review and approval before incorporating them into the HNP final safety analysis report (FSAR) and RNP UFSAR. Provide appropriate documentation for the RETRAN-3D and VIPRE-01 models according to SRP 15.0.2,Section III.2.A. The documentation should provide guidance for selecting or calculating all input parameters and code options and should also include transient-and accident-specific modeling guidelines.
Duke Energy RAI 1 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P presents methods for the applicable non-loss-of-coolant accident events described in Sections 15.1 to 15.6 of the HNP FSAR and RNP UFSAR. DPC-NE-3009-P discusses simulation codes and models, code options, safety analysis physics parameters, and transient-specific input assumptions and models.
Upon NRC approval, DPC-NE-3008-P and DPC-NE-3009-P will be added to the Technical Specifications for HNP and RNP. DPC-NE-3008-P and DPC-NE-3009-P will be used in thermal-hydraulic transient analyses as a portion of the overall Duke Energy methodology for HNP/RNP
RA-16-0034 Page 2 of 65 cycle reload safety analyses. There are additional methodology reports and analyses related to the cycle reload safety analyses. These other reports have been submitted to the NRC Staff for review and approval. The appropriate HNP FSAR and RNP UFSAR changes will be processed once core designs using these reports are implemented.
RA-16-0034 Page 3 of 65 NRC RAI 2 When discussing the RETRAN-3D models for HNP and RNP, Section 4.1 of DPC-NE-3008-P states that, "Heat conductors are also modeled using similar detail as in Tables 3.2-1 and 3.2-2 of DPC-NE-3000, with various changes... " This is the only information provided in Duke Energy's LAR regarding the heat conductor modeling for the HNP and RNP RETRAN-3D models.
SRP 15.0.2,Section III.2.B, states that, "For changes to previously approved models, the reviewers can limit their review to the new material if they determine there is nothing new that will invalidate the previous approval." However, stating that heat conductors are modeled "using similar detail" as the previous model, with "various changes," does not provide sufficient detail for the NRC staff to understand what has changed. Provide additional detail about the heat conductor modeling to justify that it meets the conditions and limitations of the RETRAN-3D safety evaluation and is capable of representing the HNP and RNP FSAR/UFSAR Chapter 15 transients.
Duke Energy RAI 2 Response DPC-NE-3000, Table 3.2-1, describes the heat conductor modeling in the RETRAN-02 Base Model for CNS/MNS with preheating steam generators. DPC-NE-3000, Table 3.2-2, lists the updates to DPC-NE-3000, Table 3.2-1, for CNS/MNS with feedring steam generators. Tables RAI-2-1 and RAI-2-2 describe the heat conductor modeling in the RETRAN-3D Base Models for HNP and RNP, respectively. In Tables RAI-2-1 and RAI-2-2, X denotes a set of volumes for Loops 1, 2 and 3, and Z denotes a set of heat conductors for Loops 1, 2 and 3.
The modeling approach and detail for HNP/RNP are generally consistent with those used to develop the approved models for CNS/MNS. [
]a,c The
RA-16-0034 Page 4 of 65 modeling for all four plants is considered to be acceptable for use in analyzing (U)FSAR Chapter 15 non-LOCA transients.
DPC-NE-3008, Section 4.2.17, describes an assessment of compliance with the conditions and limitations in the NRCs generic Safety Evaluation Report (SER) for the RETRAN-3D code. RAI 13 requests additional information regarding conditions and limitations not explicitly addressed in DPC-NE-3008. As discussed in the response to RAI 13, the additional information supports the previous assessment of compliance with the NRCs generic SER on RETRAN-3D.
Minor changes to Tables RAI-2-1 and/or RAI-2-2 are anticipated to improve consistency between HNP and RNP. For example, heat conductors for HNP may be renumbered for better correspondence between the heat conductor and volume numbers. Adjustments will also be required for selected modeling features of the (U)FSAR Chapter 15 transient analysis methodology described in DPC-NE-3009. For example, dividing the reactor vessel into two azimuthal regions as described in Section 3.2 of DPC-NE-3009 will necessitate corresponding adjustments to the heat conductor modeling. These changes are noted for completeness and do not affect the general statement of the adequacy of the heat conductor modeling.
RA-16-0034 Page 5 of 65 Table RAI-2 HNP RETRAN-3D Base Model Heat Conductors Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 6 of 65 Table RAI-2 HNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 7 of 65 Table RAI-2 HNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 8 of 65 Table RAI-2 HNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 9 of 65 Table RAI-2 RNP RETRAN-3D Base Model Heat Conductors Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 10 of 65 Table RAI-2 RNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 11 of 65 Table RAI-2 RNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 12 of 65 Table RAI-2 RNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 13 of 65 Table RAI-2 RNP RETRAN-3D Base Model Heat Conductors (Continued)
Conductor(s)
Volume(s)
Description Material a,c
RA-16-0034 Page 14 of 65 NRC RAI 3 The control systems used in the RETRAN-3D models for HNP and RNP are not discussed in DPC-NE-3008-P. It is unclear if the control systems are being modeled with an appropriate level of fidelity (see SRP 15.0.2,Section III.2.B). Provide an overview of the control system modeling.
This overview should include a discussion of the valve modeling, pressure relief valve setpoints, delay times, reactor trip setpoints, and so forth. If the discussion does not provide each parameter value, it should at least include a discussion of the basis used to develop each value to be used in future analyses of record.
Duke Energy RAI 3 Response The overview provided below describes how RETRAN-3D control systems, trips, and trip functions are used in the HNP/RNP base models to model the following:
Process Variable Indications: RETRAN control systems are used to convert thermal-hydraulic conditions into the form in which they are output by plant instrumentation.
Examples of these control systems include pressurizer pressure, pressurizer level, RCS loop temperature, steam line pressure, and SG level.
Reactor Protection System Functions: Examples of modeled reactor protection system functions include 1) over-temperature T, 2) over-power T, 3) low pressurizer pressure, 4) low-low SG narrow range liquid level, and 5) high neutron flux.
Engineered Safeguards Functions: Examples of modeled ESFAS functions include 1)
ECCS pump start on safety injection, 2) steam line isolation on high steam flow concurrent with low steam line pressure, and 3) AFW pump start on low-low SG narrow range level.
Plant Control Systems: The general approach for modeling plant control systems is the same as DPC-NE-3000-P: control systems are modeled based on their relative importance to the transient being evaluated. While less-detailed models are adequate for some control systems, the modeling of other plant control systems such as pressurizer pressure control
RA-16-0034 Page 15 of 65 more closely matches the actual plant controller. For example, the RETRAN-3D model of pressurizer pressure control uses a proportional plus integral controller, which emulates the actual plant controller.
Transient Boundary Conditions: Examples of modeled transient boundary conditions include 1) adjusting reactivity via a scram curve; 2) modulating valves and control trips according to their setpoints and delays (refer to RAI-16); or 3) adjusting positive or negative fill flow rates.
Component response times, setpoints, and capacities supported in the accident analyses are presented in FSAR, Tables 15.0.3-5 and 15.0.6-2 (HNP) and UFSAR, Tables 15.0.7-1 and 15.0.8-1 (RNP). DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of the HNP/RNP RETRAN-3D base models in completing system transient analyses for the HNP FSAR and RNP UFSAR. Section 5.0 of DPC-NE-3009-P states that the performance characteristics of these control systems, trips, and trip functions are modeled to give conservative response and provide bounding consequences for each of the events.
RA-16-0034 Page 16 of 65 NRC RAI 4 Use of the Chexal-Lellouche correlation is the subject of several conditions and limitations of the NRC staff's safety evaluation on RETRAN-3D. Duke Energy indicated in Section 4.2.4 of DPC-NE-3008-P that the Chexal-Lellouche drift flux correlation will be used, and provided the NRC staff's approval of the correlation at the Oconee as a basis to judge it acceptable in the present application. Considering that the application of the Chexal-Lellouche correlation for HNP and RNP is different from the previous application at Oconee due to design differences between the plants, describe how Duke Energy ensures that the correlation will be used within its range of applicability at HNP and RNP as required by the NRC's RETRAN-3D safety evaluation.
Duke Energy RAI 4 Response DPC-NE-3008-P, Section 4.2.4, states that the HNP/RNP RETRAN-3D Base Models use the Algebraic Slip Model with the Chexal-Lellouche correlation in [
]a,c. According to Reference RAI-4-1, Table 5-1, the Chexal-Lellouche steam-water database covers wide ranges of void fraction, mass flux, pressure, heat flux and subcooling. [
]a,c According to Reference RAI-4-1, Table 5-1, the Chexal-Lellouche steam-water database also covers a wide range of hydraulic diameter. [
]a,c According to the Staff Position on RETRAN-3D SER Condition 16, results of analyses using the Chexal-Lellouche correlation in the pressure ranges of 12 to 17 MPa (1,740 to 2,466 psia) and
RA-16-0034 Page 17 of 65 7.5 to 10 MPa (1,088 to 1,450 psia) must be carefully reviewed. [
]a,c According to the Staff Position on RETRAN-3D SER Condition 16, the Chexal-Lellouche correlation cannot be used in situations where CCFL is important unless validation for appropriate geometry and expected flow conditions is provided. [
]a,c References RAI-4-1 Chexal, B., et al., Void Fraction Technology for Design and Analysis, TR-106326, March 1997.
RA-16-0034 Page 18 of 65 NRC RAI 5 RETRAN-3D has a non-equilibrium volume option that allows the liquid and vapor regions in a given volume to have different temperatures. Duke Energy stated in Section 4.2.4 of DPC-NE-3008-P that the non-equilibrium volume option is used in the pressurizer volume. This is a departure from the previously-approved model described in DPC-NE-3000-PA, where it was also applied in the reactor vessel head. However, Duke Energy stated that licensing applications of the RETRAN-3D model ((
)). Clarify whether such applications will receive further NRC review. If not, in accordance with the guidance provided in SRP 15.0.2 and Regulatory Guide 1.203, justify how Duke Energy will determine ((
)).
Duke Energy RAI 5 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P has been submitted for NRC Staff review. Section 5.0 of DPC-NE-3009-P discusses the use of the RETRAN-3D Two-Region Non-Equilibrium Volume Model.
RA-16-0034 Page 19 of 65 NRC RAI 6 Duke Energy stated in Section 4.2.4 of DPC-NE-3008-P that the RETRAN-3D models for HNP and RNP do not model ((
)). This is a change from the previously-approved DPC-NE-3000-PA methodology. SRP 15.0.2,Section III.5, provides review guidance to the NRC staff regarding the conservatism of changes to previously-approved methodologies. ((
)). Provide justification why this modeling choice is acceptable, given that it is a departure from the previously-approved DPC-NE-3000-PA methodology.
Duke Energy RAI 6 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P has been submitted for NRC Staff review. Section 5.0 of DPC-NE-3009-P provides the modeling choice justification regarding the RETRAN-3D Inter-Region Heat Transfer Model.
RA-16-0034 Page 20 of 65 NRC RAI 7 RETRAN-3D has a non-conducting heat exchanger model that allows for simple heat transfer to or from fluid volumes without using a heat conductor. In Section 4.2.5 of DPC-NE-3008-P, Duke Energy stated that, Licensing applications of the RETRAN-3D models for HNP and RNP may incorporate other uses of non-conducting heat exchangers to model, for example, ambient heat losses. Clarify whether such applications will receive further NRC review. If not, in accordance with the guidance provided in SRP 15.0.2 and Regulatory Guide 1.203, provide additional information on the uses of non-conducting heat exchangers that will be performed under this methodology, including specification of analysis scenarios and relevant modeling assumptions, and justify why the non-conducting heat exchanger model is appropriate for the scenarios modeled.
Duke Energy RAI 7 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P has been submitted for NRC Staff review. Section 5.0 of DPC-NE-3009-P discusses the use of the RETRAN-3D Non-Conducting Heat Exchanger Model.
RA-16-0034 Page 21 of 65 NRC RAI 8 RETRAN-3D has a local conditions heat transfer model that allows the heat transfer from a heat conductor to vary depending on local fluid conditions in the attached fluid volume. In Section 4.2.6 of DPC-NE-3008-P, Duke Energy stated that, Licensing applications of the HNP and RNP RETRAN models may use the local conditions heat transfer model for other volumes, such as the reactor vessel head, when conditions warrant. Clarify whether such applications will receive further NRC review. If not, In accordance with the guidance provided in SRP 15.0.2 and Regulatory Guide 1.203, provide additional information on the uses of the local conditions heat transfer model that will be performed under this methodology, including specification of analysis scenarios and relevant modeling assumptions. Discuss how Duke Energy will determine that conditions in the transient scenario warrant the use of the model.
Duke Energy RAI 8 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P has been submitted for NRC Staff review. Section 5.0 of DPC-NE-3009-P discusses the use of the RETRAN-3D Local Conditions Heat Transfer Model.
RA-16-0034 Page 22 of 65 NRC RAI 9 In Section 4.2.7 of DPC-NE-3008-P, Duke Energy states that the steady-state initialization process used for HNP and RNP is similar to the process used for the McGuire Nuclear Station (MNS) and the Catawba Nuclear Station (CNS) in terms of the inputs that may be adjusted.
- a. In order for the NRC staff to determine the scope of this aspect of the review, specify what has changed relative to the MNS/CNS process.
- b. How does Duke Energy ensure that these adjustments to initialize the steady-state model do not non-conservatively impact subsequent transient analyses?
Duke Energy RAI 9 Response Response to RAI-09a The RETRAN-3D steady-state initialization process is used to obtain stable initial conditions for transient simulations. The RETRAN-3D Users Manual (Reference RAI-9-1) presents various methods that were not available during the development of the RETRAN-02 MNS/CNS method, such as the RETRAN-3D Low Power Steam Generator Steady-State Initialization (approved by NRC Staff position on SER Condition 29) and the RETRAN-3D Off-Rated Power Initialization Procedure. The newer methods perform automatic calculations to obtain the desired initial conditions that would otherwise require multiple manual calculations. Regardless of the method selected, some physical parameter must be adjusted (e.g., SG pressure, SG tube bundle inlet enthalpy, or SG tube conductivity) to achieve the overall primary-to-secondary energy balance and the desired initial conditions.
The following overview describes the general approach to obtaining stable initial conditions for each RETRAN-3D transient analysis using the HNP/RNP base models. This overview generally follows Section 3.2.7.1 of DPC-NE-3000-PA, which describes the process used for MNS/CNS analyses.
Primary System Conditions: For the most part, the primary system steady-state initialization process for HNP/RNP is consistent with the MNS/CNS process. The main difference is
RA-16-0034 Page 23 of 65 specifying the enthalpy in the reactor vessel lower plenum (rather than in the cold leg as in the MNS/CNS process). For HNP/RNP, primary system conditions are set by specifying core power, pressurizer pressure, pressurizer level, flow rates, and reactor vessel lower plenum enthalpy.
In parallel flow paths, such as the core and core bypass, a junction form loss coefficient may be unspecified in the input. With the flows specified, the RETRAN-3D initialization option adjusts the unspecified form loss coefficient to balance the pressure losses through the parallel paths.
Secondary System Conditions: For the most part, the secondary system steady-state initialization process for HNP/RNP is consistent with the MNS/CNS process. The main difference pertains to the use of the bubble rise model in the separator; a mixture level is set in the separator for the HNP/RNP base models. Section 4.1.2.2 of DPC-NE-3008-P provides additional information related to steam generator modeling in the HNP/RNP base models. The secondary system conditions are set by specifying MFW and main steam flow, MFW enthalpy, upper downcomer enthalpy, circulation flows, steam dome pressure, and levels in the steam dome and separator.
To obtain the desired initial SG level and mass, circulation flows, junction form loss coefficients, and mixture levels in the downcomer and separator may be adjusted. Bubble rise parameters may also be adjusted. During a steady-state initialization, the RETRAN initialization routine adjusts the SG tube surface area such that the heat transfer through the SG tubes matches the desired heat transfer. To minimize this RETRAN-calculated adjustment, the SG tube thermal conductivity, SG tube bundle inlet enthalpy, or SG steam dome pressure may be adjusted manually or automatically depending on the method chosen.
Response to RAI-09b The conservatism in the transient analysis is driven primarily by conservative analysis assumptions related to initial conditions, system and component availability, etc. It is incumbent on the user to consider the analysis objective (e.g., an evaluation of primary system over-pressure or a short-term analysis of core cooling capability); select the appropriate parameter
RA-16-0034 Page 24 of 65 for adjustment in performing the steady state initialization; and ensure the adjustments do not have an adverse effect on analysis results.
RETRAN-3D SER Condition 29 discusses the potential for ill-posed steady-state information to affect the transient solution and refers to various techniques to resolve this issue, such as output checks (the RETRAN-3D calculated adjustments, for example) and the use of null transients. These techniques are retained for applications of the RETRAN-3D base models for HNP and RNP.
References RAI-9-1 RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 3: Users Manual, NP-7450(A), Vol. 3, Rev. 10, September 2014.
RA-16-0034 Page 25 of 65 NRC RAI 10 SRP 15.0.2,Section II, states that chosen mathematical models and the solution of those models must be able to predict the important physical phenomena reasonably well from both qualitative and quantitative points of view. While the option for using a donor-cell formulation for computing momentum flux was removed from RETRAN-3D, the built-in arithmetic averaging formulation is known to produce numerically unstable results in certain situations. How will Duke Energy avoid these instabilities, and, given that the donor-cell formulation has been removed, what tools are available to correct them if they arise?
Duke Energy RAI 10 Response Duke Energy has two ongoing processes that help minimize the potential for being significantly impacted by numerical instabilities such as the one described. The first process involves various interactions with the vendor responsible to maintain the RETRAN-3D code, the Computer Simulation and Analysis group of Zachry Nuclear Engineering, Inc. (CSA/ZNE). Examples of these interactions include routine usage of the Computer Code Manuals (theory, input guidelines, etc.); periodic evaluation of Software Problem Reports (using the Duke Energy Corrective Action Program); regular participation in User Group meetings; and focused requests for expert consultation. The second process involves careful review of the analysis results. This process is an essential part of applying any system thermal-hydraulic computer code and leverages many years of experience in applying the RETRAN-02 and RETRAN-3D codes to the Catawba, McGuire and Oconee Nuclear Stations.
The RAI statement suggests a scenario whereby an analysis is significantly impacted by numerical instabilities that are attributed to the arithmetic-averaging formulation for computing momentum flux. Where adequate and justified, the problem may be resolved by deactivating the momentum flux terms in the affected junction(s). Otherwise, Duke Energy would consult with CSA/ZNE to identify options for resolving the observed behavior. If the recommended solution were to reactivate the donor-cell formulation for computing momentum flux, then Duke Energy would consult with CSA/ZNE to implement the required updates.
RA-16-0034 Page 26 of 65 NRC RAI 11 RETRAN-3D has a general transport model to calculate the transport and concentration of chemicals such as boric acid. In Section 4.2.15 of DPC-NE-3008-P, Duke Energy stated that, Although the RETRAN-3D base models for HNP and RNP do not use the general transport model, it may be used for licensing applications of the HNP and RNP RETRAN models where significant reactivity effects associated with boron transport are encountered. Clarify whether such applications will receive further NRC review. If not, in accordance with the guidance provided in SRP 15.0.2 and Regulatory Guide 1.203, provide additional information on the uses of the boron transport model that will be performed under this methodology, including specification of analysis scenarios and relevant modeling assumptions, and discuss how Duke Energy will determine that reactivity effects associated with boron transport are significant.
Duke Energy RAI 11 Response DPC-NE-3008-P describes RETRAN-3D and VIPRE-01 base models and assumptions for HNP and RNP and qualifies these models for licensing applications. DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. DPC-NE-3009-P has been submitted for NRC Staff review. Section 5.0 of DPC-NE-3009-P discusses the use of the RETRAN-3D General Transport Model.
RA-16-0034 Page 27 of 65 NRC RAI 12 Duke Energy stated in Section 4.2.16 of DPC-NE-3008-P that the RETRAN-3D accumulator component model is described and validated in Section III.11.0 of Electric Power Research Institute (EPRI) NP-7450, Volume 4. However, this section does not exist in the version of the RETRAN-3D assessment manual available from EPRI. Provide a description and assessment of the accumulator model.
Duke Energy RAI 12 Response The document titled RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 4: Applications Manual, NP-7450(A), Vol. 4, Rev. 9, is provided in Attachment 5.Section III.11.0 of this document describes the tests performed to qualify the pressure and level response of the RETRAN-3D accumulator model.
RA-16-0034 Page 28 of 65 NRC RAI 13 Duke Energy evaluates the RETRAN-3D safety evaluation report (SER) conditions and limitations in Section 4.2.17 of DPC-NE-3008-P. However, this section lists only several conditions from the NRC staffs safety evaluation of RETRAN-3D. It is implied that disposition of the conditions and limitations not listed in Section 4.2.17 of DPC-NE-3008 can be found in Appendix C to DPC-NE-3000-PA, but this is not clear. Provide dispositions to the RETRAN-3D conditions and limitations not discussed in DPC-NE-3008-P, or confirm that they are provided in DPC-NE-3000-PA, and justify that the prior disposition remains applicable.
Duke Energy RAI 13 Response As part of the review of RETRAN-3D, the NRC Staff examined the limitations and conditions of use of earlier RETRAN versions to determine which are still applicable and which have been resolved by new models and additions in RETRAN-3D (Reference RAI-13-1). Appendix C of DPC-NE-3000-PA, Revision 5a (hereafter DPC-NE-3000), provides an assessment of the limitations and conditions of use from the NRCs generic Safety Evaluation Report (SER) on the RETRAN-3D computer code for the Oconee RETRAN-3D model (Reference RAI-13-2). Section 4.2.17 of DPC-NE-3008-P provides a HNP or RNP-specific evaluation of these limitations and conditions of use. A supplemental discussion is provided below for limitations and conditions of use considered not applicable or for which the NRC staff or the Duke Energy position in Appendix C of DPC-NE-3000 apply.
Condition 1:
This limitation or condition is not applicable to the HNP/RNP RETRAN-3D base models. Section 4.2.1 of DPC-NE-3008-P states that the HNP/RNP RETRAN-3D base models use a point kinetics model.
Condition 2:
The Duke Energy position described in DPC-NE-3000, page C-4, is retained for the HNP/RNP base models.
Condition 3:
The Duke Energy position described in DPC-NE-3000, page C-4, is retained for the HNP/RNP base models.
RA-16-0034 Page 29 of 65 Condition 4:
This limitation or condition is not applicable to the HNP/RNP RETRAN-3D base models. Section 4.2.1 of DPC-NE-3008-P states that the HNP/RNP RETRAN-3D base models use a point kinetics model.
Condition 5:
This limitation or condition is not applicable because Duke Energy does not use the metal-water heat generation model.
Condition 6:
This limitation or condition is not applicable because Duke Energy does not use the RETRAN-3D five-equation constrained non-equilibrium model for the thermal-hydraulics field equations.
Condition 7:
The Duke Energy position described in DPC-NE-3000, page C-5, is retained for the HNP/RNP base models.
Condition 8:
This limitation or condition is not applicable because HNP and RNP are not BWRs.
Condition 9:
This limitation or condition is not applicable because HNP and RNP are not BWRs.
Condition 10: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models use algebraic slip instead of dynamic slip.
Condition 11: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models do not use the optional linear gap thermal expansion model.
Condition 12: This concern is not applicable because the HNP/RNP RETRAN-3D base models do not use the noncondensable gas flow model.
Condition 13: The Duke Energy position described in DPC-NE-3000, page C-7, is retained for the HNP/RNP base models.
Condition 14: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P.
Condition 15: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-7).
Condition 16: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P. Supplementary information is provided in the response to RAI-4 and Section 3.2 of DPC-NE-3009-P.
Condition 17: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models use algebraic slip instead of dynamic slip.
Condition 18: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P. The modeling practices in the NRC Staff position are noted. Supplementary information
RA-16-0034 Page 30 of 65 regarding thermal stratification in the pressurizer is provided in the response to RAI-21.
Condition 19: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models do not use the RETRAN-3D separator model.
Condition 20: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P.
Condition 21: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models do not model BWR jet pumps.
Condition 22: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models do not use the turbine model.
Condition 23: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-10).
Condition 24: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P.
Condition 25: The RETRAN-3D Temperature Transport Delay Option is not used in the HNP/RNP RETRAN-3D base models (DPC-NE-3008, Section 4.2.10). If the RETRAN-3D Temperature Transport Delay Option is used, the usage will be consistent Duke Energy position described in DPC-NE-3000, pages C-10 and C-
- 11.
Condition 26: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models do not use the RETRAN-3D stand-alone DNBR model.
Condition 27: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-11).
Condition 28: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P.
Condition 29: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-12).
Supplementary information is provided in the response to RAI-9.
Condition 30: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-12).
Supplementary information is provided in the response to RAI-4 and Section 3.2 of DPC-NE-3009-P.
Condition 31: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-12).
Condition 32: This limitation or condition is not applicable to the HNP/RNP RETRAN-3D base models. Section 4.2.1 of DPC-NE-3008-P states that the HNP/RNP RETRAN-3D base models use a point kinetics model.
Condition 33: The Duke Energy position described in DPC-NE-3000, page C-13, is retained for the HNP/RNP base models.
RA-16-0034 Page 31 of 65 Condition 34: RETRAN-3D has the same model as RETRAN-02 and subsequent versions that have been approved for use. The boron tracking model is active in the HNP/RNP RETRAN-3D base models, but the associated reactivity effects are not modeled in the base models. Supplementary information is provided in Section 3.2 of DPC-NE-3009-P describing the application of this model for specific transients.
Condition 35: The application of the HNP/RNP RETRAN-3D base models in transients where mixing and core flow are important is addressed in Section 3.2 of DPC-NE-3009-P.
Condition 36: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-13).
Condition 37: This concern is resolved per the NRC Staff position (DPC-NE-3000, p. C-13).
Condition 38: The RETRAN-3D algebraic slip model is not used in primary side junctions in the HNP/RNP base models. Voiding may occur on the primary side during a main steam line break. This is addressed in Section 3.2 of DPC-NE-3009-P.
Condition 39: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models are not used in BWR transient analysis.
Condition 40: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P. Supplementary information is provided in the response to RAI-14.
Condition 41: This limitation or condition is not applicable because the HNP/RNP RETRAN-3D base models are not used in BWR transient analysis.
Condition 42: This limitation or condition is not applicable because Duke Energy does not use the RETRAN-3D five-equation constrained non-equilibrium model for the thermal-hydraulics field equations.
Condition 43: The application of the HNP/RNP RETRAN-3D base models for transient analysis is addressed in Section 3.2 of DPC-NE-3009-P.
Condition 44: The Duke Energy position described in DPC-NE-3000, pages C-15 and C-16, is retained for the HNP/RNP base models. It is noted that Volumes 3 and 5 of the EPRI RETRAN-3D documentation contain a significant amount of user guidelines regarding modeling option selection, in particular for the new RETRAN-3D models and options. The following documents are provided in Attachment 5:
RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 1: Theory and Numerics Manual, NP-7450(A), Vol. 1, Rev. 10.
RA-16-0034 Page 32 of 65 RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 2: Programmers Manual, NP-7450(A), Vol.
2, Rev. 10.
RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 3: Users Manual, NP-7450(A), Vol. 3, Rev.
- 10.
RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 4: Applications Manual, NP-7450(A), Vol. 4, Rev. 9.
RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Systems; Volume 5: Modeling Guidelines, NP-7450(A), Vol.
- 5.
Condition 45: This condition is addressed in Section 4.2.17 of DPC-NE-3008-P. Section 3.2 of DPC-NE-3009-P provides supplementary information related to the application of these models. Duke is not proposing to use these models for best-estimate licensing applications.
References RAI-13-1 NRC letter, Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems (TAC No. MA4311), dated January 25, 2001 (ADAMS Accession No. ML010470342).
RAI-13-2 DPC-NE-3000-P-A, Thermal-Hydraulic Transient Analysis Methodology, Rev. 5a, October 2012.
RA-16-0034 Page 33 of 65 NRC RAI 14 Duke Energys disposition of RETRAN-3D SER condition 40 states that several new control block models are used, which have not been previously reviewed by the NRC staff, yet does not provide any detail beyond naming the new blocks used in the model. Provide additional detail on the function and purpose of each of these control blocks (or a citation to a publicly-available document where such information may be found) and how they are used in the HNP and RNP RETRAN-3D models, in order to comply with RETRAN-3D SER condition 40.
Duke Energy RAI 14 Response The disposition of RETRAN-3D SER Condition 40 lists three new control blocks that are used in the base models for HNP and RNP. These control blocks allow selected operations on a series of minor edits and are denoted as super blocks. The control blocks are named SSM, SMX, and SMN for super summer, super maximum, and super minimum, respectively.
The new control blocks simplify the input that would be required to apply the SUM, MAX, or MIN to multiple inputs. For example, using the original SUM block, the total liquid mass (LIQM) of Volumes 1, 2, 3, and 4 could be calculated as follows:
LIQM 1 LIQM 2 LIQM 3 LIQM 4 SUM SUM SUM LIQM 1 + LIQM 2 + LIQM 3 + LIQM 4 Using the new SSM block simplifies the logic to the following:
LIQM 1, 2, 3, 4 SSM LIQM 1 + LIQM 2 + LIQM 3 + LIQM 4
RA-16-0034 Page 34 of 65 In the RETRAN-3D base models for HNP and RNP, examples of applying the new blocks are to calculate the minimum and maximum reactor vessel average temperature for Loops 1, 2, and 3 (SMN and SMX) and the total secondary-side liquid mass in each steam generator (SSM).
RA-16-0034 Page 35 of 65 NRC RAI 15 SRP 15.1.2 states that conservative scram characteristics should be assumed when analyzing a decrease in feedwater flow. However, the rod insertion speed and total negative reactivity insertion resulting from a reactor trip in the HNP and RNP RETRAN-3D models are unclear from the documentation provided in DPC-NE-3008-P. Please provide trip reactivity curves for the HNP and RNP models.
Duke Energy RAI 15 Response The rate and magnitude of the negative reactivity insertion following a reactor trip is dependent upon the control rod position versus time and the variation in reactivity worth as a function of rod position. [
]a,c The data is normalized in both the rod position and time domains to produce a normalized rod position versus normalized time curve. The reference point for insertion is the top of the dashpot. This position represents approximately 80% of the total rod insertion travel into the active fuel region, and is chosen because the Technical Specification required rod drop timing test acceptance criterion is based on the time when the reactor trip breakers are opened until the time the control rod reaches the top of the dashpot. The rod position versus time curve assumed in the Safety Analysis is determined by multiplying the normalized curve by the Technical Specification rod drop time acceptance criterion.
The position dependent insertion of negative reactivity following a reactor trip is calculated at HFP assuming the highest worth stuck rod remains in its fully withdrawn position. The calculation is performed assuming a bottom peaked power distribution to conservatively delay the amount of negative reactivity inserted following initial control rod movement. The shape of the curve is variable and is dependent upon plant axial flux difference operating limits. Example normalized reactivity worth versus normalized rod position curves for Harris and Robinson are shown in Figures RAI-15-1 and RAI-15-2. These curves are typically referred to as trip reactivity shape curves. The trip shape curve is multiplied by the minimum trip worth to obtain trip reactivity versus rod position.
RA-16-0034 Page 36 of 65 Minimum trip reactivity is the total available rod worth that can be inserted into the reactor core following a reactor trip. It is the inserted worth from all control rods reduced by allowances for the highest worth stuck rod, a 10% rod worth uncertainty (applied to the N-1 worth), and a factor to account for the control rods potentially being at their Technical Specification rod insertion limit. The remaining worth is the minimum trip reactivity. The trip reactivity shape curve, the normalized rod position versus time curve and the minimum trip reactivity, are combined to calculate the minimum trip reactivity versus normalized rod position curve assumed in the Safety Analysis.
The minimum trip reactivity versus time curves assumed in the Harris and Robinson demonstration analyses were based on the analysis of record curves for consistency in the benchmarks. Example curves developed using the above method are shown in Figures RAI 3 and RAI-15-4 for Harris and Robinson, respectively. Both the analysis of record curves and the curves presented in Figures RAI-15-3 and RAI-15-4 conservatively delay the insertion of negative reactivity following reactor trip. Conservatisms included in the overall trip reactivity modeling include:
allowances for instrumentation delay and uncertainty from the time the trip condition is reached until the start of rod movement control rod drop times that bound measured times delay in the negative reactivity inserted following initial control rod movement reduction in the trip reactivity worth Cycle-specific reload checks are performed to verify minimum trip reactivity versus rod position assumption. Plant tests are performed prior to reactor startup following each refueling outage to verify that rod drop times are within the rod insertion analysis assumption, and to verify that measured control rod worths are within 10% of predicted. Together, these verifications confirm the acceptability of the negative reactivity versus time assumption employed in the Safety Analysis.
RA-16-0034 Page 37of65 Figure RAl-15-1 Example Harris Trip Reactivity Shape (Normalized Worth vs. Normalized Rod Position) a,c RA-16-0034 Page 38of65 Figure RAl-15-2 Example Robinson Trip Reactivity Shape (Normalized Worth vs. Normalized Rod Position) a,c RA-16-0034 Page 39of65 Figure RAl-15-3 Example Harris Normalized Worth Versus Time Figure RAl-15-4 Example Robinson Normalized Worth Versus Time Curve a,c a,c
RA-16-0034 Page 40 of 65 NRC RAI 16 Section 4.3.2 of DPC-NE-3008-P compares results of the RETRAN-3D model for HNP to the AOR for the plant for the turbine trip event. For events like the turbine trip that result in a decrease in heat removal by the secondary system, SRP 15.2.3,Section III.6.J, instructs the NRC staff to review pressure safety and relief valve flow rates. Duke Energy states in Section 4.3.2 that the, RETRAN-3D models of the pressurizer safety valves and main steam safety valves were justified by comparing the valve flows with the results documented in the AOR.
The AOR does not present valve flows, and valve flow rates from the RETRAN-3D analysis were not provided. Given that the pressurizer safety valve and main steam safety valve flows are extremely important to the overall system response, how does Duke Energy justify that the valve flows in the RETRAN-3D models are reasonable? Are there other parameters that were compared to the AOR? How are these valves modeled?
Duke Energy RAI 16 Response Section 4.3.2 of DPC-NE-3008-P states that the [
] a,c In the transient analysis comparison presented in Section 4.3.2, the RETRAN-3D modeling of the opening and closing characteristics of the valves [
]a,c With the importance of the pressurizer safety valve and main steam safety valve flows to this transient, the good agreement in the overall system response between the RETRAN-3D analysis and the AOR indicates that the RETRAN-3D models for these valves are reasonable.
DPC-NE-3009-P supplements DPC-NE-3008-P and describes the application of these models in completing system transient analyses for the HNP FSAR and RNP UFSAR. Section 5.0 of
RA-16-0034 Page 41 of 65 DPC-NE-3009-P states that the performance characteristics of these control systems, trips, and trip functions are modeled to give conservative response and provide bounding consequences for each of the events.
RA-16-0034 Page 42 of 65 NRC RAI 17 Section 4.3.3 of DPC-NE-3008-P compares results of the RETRAN-3D model for HNP to the AOR for the plant for the main feedwater line break. For this event, SRP 15.2.8,Section I.1.H, identifies steam generator water level as a parameter of importance.
In Figure 4.3-22, there is a difference between the steam generator level in the unaffected steam generators in the RETRAN-3D analysis relative to the AOR, starting around 5 seconds.
Explain the cause of (a) the decrease and subsequent sudden increase in steam generator level between approximately 5 and 10 seconds and (b) the holdup of steam generator level between approximately 20 and 30 seconds.
Duke Energy RAI 17 Response The overall steam generator level predictions from the RETRAN-3D analysis agree reasonably well with the AOR. Between 5 and 30 seconds, the total liquid mass in each steam generator decreases steadily (Figure RAI-17-1). The intact steam generator level does not actuate any trips during the event. Therefore, the intact steam generator level behaviors observed between 5 to 10 seconds and 20 to 30 seconds have minimal impact on the overall transient response.
The decrease in the intact steam generator levels starting around 5 seconds is attributed to
[
]a,c The sudden increase in the intact steam generator levels starting around 9 seconds is attributed to [
RA-16-0034 Page 43 of 65
]a,c The holdup of the intact steam generator levels between approximately 20 and 30 seconds is attributed to an increase in [
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0 5
10 15 20 25 30 35 Time (s)
RA-16-0034 Page45 of65 Figure RAl-17-2 HNP FWLB Event - no LOOP -RETRAN-30 Loop 2 Steam Generator Liquid Mass Distribution a,c
RA-16-0034 Page 46 of 65 NRC RAI 18 Section 4.3.4 of DPC-NE-3008-P compares results of the RETRAN-3D model for RNP to the AOR for the plant for the loss of normal feedwater flow transient. SRP 15.2.7 Section III identifies coolant conditions as a key parameter for this event. Figures 4.3-36 through 4.3-38 show the reactor vessel inlet, outlet, and average temperatures for the RETRAN-3D analysis and the AOR. In these plots, the vessel outlet and average temperatures in the AOR rise slightly after the initial drop, before decreasing again over time. The RETRAN-3D analysis does not show this same phenomenon. Explain the difference in the phenomenon. Discuss and justify that the RETRAN-3D analysis is modeled appropriately.
Duke Energy RAI 18 Response The loss of normal feedwater event is defined to result from a complete loss of the normal feedwater flow to all the steam generators. The loss of normal feedwater results in an immediate reduction of steam generator liquid level and a decrease in primary to secondary heat transfer. Section 4.3.4 of DPC-NE-3008-P compares results of the RETRAN-3D model for RNP to the analysis of record for this event.
The curve labeled RETRAN-3D Vessel Inlet in Figure 4.3-36 of DPC-NE-3008 represents the fluid temperature in RETRAN-3D [
]a,c. The curve labeled RETRAN-3D Vessel Outlet in Figure 4.3-38 of DPC-NE-3008 represents the fluid temperature in RETRAN-3D
[
]a,c. As shown in Figure 4.1-1 of DPC-NE-3008, [
]a,c The curve labeled RETRAN-3D Vessel Average in Figure 4.3-37 of DPC-NE-3008 represents [
]a,c The RETRAN-3D results presented in Figures 4.3-36 to 4.3-38 of DPC-NE-3008 are re-plotted for clarity in Figure RAI-18-1. The re-plotted figure shows that the RETRAN-3D results exhibit similar behavior to the AOR: the vessel outlet and average temperatures rise slightly after the initial drop, before decreasing again over time. The temperature rise in the RETRAN-3D vessel outlet temperature is approximately 1 °F, which is similar to the temperature rise in the AOR.
RA-16-0034 Page 47 of 65 Additional discussion of the temperature rise is provided below to show that the RETRAN-3D analysis is modeling this behavior appropriately.
Figure RAI-18-2 presents the transient results for the Reactor Coolant System (RCS) temperature difference between the inlet and outlet of the steam generator. The results are presented for RCS Loop 3, but the Loops 1 and 2 results are almost identical to Loop 3. In response to the loss of normal feedwater at time zero, SG pressure decreases temporarily, which temporarily improves primary-to-secondary heat transfer: the temperature difference between the inlet and outlet of the steam generator increases. When scram rod insertion begins at 38.9 seconds (Table 4.3-10 of DPC-NE-3008), the coincident turbine trip increases SG pressure and decreases primary-to-secondary heat transfer: the temperature difference between the inlet and outlet of the steam generator decreases and the vessel inlet temperature increases a few seconds later (Figure RAI-18-1). The first stage MSSVs begin to open at 50 seconds, which temporarily improves primary-to-secondary heat transfer: the temperature difference between the inlet and outlet of the steam generator increases and the vessel inlet temperature decreases a few seconds later (Figure RAI-18-1). The subsequent opening of the second and third stage MSSVs stabilize primary-to-secondary heat transfer.
RA-16-0034 Page 48of65 C' -
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- 650 600 sso 500 Figure RAl-18-1 Reactor Coolant System Temperatures 1
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RA-16-0034 Page 50 of 65 NRC RAI 19 Section 4.3.5 of DPC-NE-3008-P compares results of the RETRAN-3D model for HNP to the AOR for the plant for the complete loss of forced reactor flow transient. SRP 15.3.2 Section III.5, identifies coolant conditions as a key parameter for this event. The discussion in the section notes that the vessel outlet temperature is higher in the RETRAN-3D analysis than it is in the AOR Figure 4.3-50 shows that the core outlet temperature is higher, even before the transient initiation in the RETRAN-3D analysis. Explain the difference between the RETRAN-3D steady-state vessel outlet temperature from the AOR and the impact on the new analysis.
Duke Energy RAI 19 Response The curve labeled RETRAN-3D T-outlet in DPC-NE-3008-P, Figure 4.3-50, represents the fluid temperature in RETRAN-3D [
]a,c. As shown in DPC-NE-3008-P, Figure 4.1-1, [
]a,c. The initial value is obtained from the RETRAN-3D output as 626.3 °F.
[
]a,c The curve labeled T-outlet in DPC-NE-3008-P, Figure 4.3-50, is based on the ANF-RELAP analysis and has an initial value of about 624 °F. [
]a,c
RA-16-0034 Page 51 of 65 NRC RAI 20 Section 4.3.6 of DPC-NE-3008-P compares results of the RETRAN-3D model for RNP to the AOR for the plant for the locked rotor event. SRP 15.3.4,Section III, instructs the NRC staff to review reactor coolant system pressure. Figure 4.3-55 shows the pressurizer and core outlet pressures for both the RETRAN-3D analysis and the AOR. Explain why the RETRAN-3D core outlet pressure curve has noticeably different trends from the AOR. Explain what causes the core exit and pressurizer pressures to drop starting between 2.5 to 3 seconds, and why the pressurizer pressure drops first in the RETRAN-3D analysis as opposed to the AOR.
Duke Energy RAI 20 Response The locked rotor event is defined to result from an instantaneous seizure of a reactor coolant pump rotor with the reactor at rated power plus uncertainty. Coolant flow in the affected loop is rapidly reduced, which results in a heatup of the primary system. Section 4.3.6 of DPC-NE-3008-P compares results of the RETRAN-3D model for RNP to the analysis of record for this event.
The curve labeled RETRAN-3D Pressurizer in Figure 4.3-55 of DPC-NE-3008 represents the pressure in RETRAN-3D [
]a,c. The curve labeled RETRAN-3D Core Outlet in 4.3-55 of DPC-NE-3008 represents the pressure in RETRAN-3D [
]a,c. As shown in Figure 4.1-1 of DPC-NE-3008, [
]a,c.
In Figure 4.3-55 of DPC-NE-3008, the RETRAN-3D and AOR results for pressurizer pressure are nearly identical through the time of minimum DNBR: 2.55 seconds in the RETRAN-3D case versus 2.25 seconds in the AOR (Table 4.3-14 of DPC-NE-3008). In the RETRAN-3D results, the pressurizer pressure trend reflects pressure relief through the pressurizer power-operated relief valve between 2.65 and 4.16 seconds. Based on available information, it is unclear when pressure relief through the power-operated relief valve occurs in the AOR. [
]a,c
RA-16-0034 Page 52 of 65 Prior to the opening of the power-operated relief valve, the differences between the RETRAN-3D and AOR results for core-exit pressure are most likely due to differences in the pressure losses in the surgeline. At steady-state with negligible pressurizer surge-line flow, the RETRAN-3D and AOR results show good agreement in the pressure difference between the core-exit and the pressurizer (Figure 4.3-55 of DPC-NE-3008). As the transient progresses and the pressurizer pressure increases during the insurge, the differences between the RETRAN-3D and AOR results become more noticeable. [
]a,c The differences between the RETRAN-3D and AOR results for pressurizer and core-exit pressure are judged to have little impact on the minimum DNBR calculation and are judged to be acceptable given the available information.
RA-16-0034 Page 53 of 65 NRC RAI 21 Section 4.3.7 of DPC-NE-3008-P compares results of the RETRAN-3D model for RNP to the AOR for the plant for uncontrolled rod control cluster assembly bank withdrawal at power event.
SRP 15.4.2,Section III.6, presents the reactor coolant system pressure as a key parameter for the event. Figure 4.3-59 shows the pressurizer pressure. Explain why the RETRAN-3D pressurizer pressure in this analysis drops substantially faster after the trip than in the AOR.
Duke Energy RAI 21 Response DPC-NE-3008-P, Table 4.3-16, shows the sequence of events for the RNP Uncontrolled RCCA Bank Withdrawal at Power event. Using the AOR (UFSAR) results as representative, RCCA insertion begins at 64.4 seconds; minimum DNBR occurs at 64.8 seconds; and maximum pressurizer pressure occurs at 65.6 seconds. DPC-NE-3008-P, Figure 4.3-59, compares the pressurizer pressure transients for the AOR and RETRAN-3D calculations. Relative to the AOR calculation, the decrease in pressurizer pressure predicted by RETRAN-3D is similar from 66 to 68 seconds; faster from 68 to 72 seconds; and slower from 72 to 79 seconds. DPC-NE-3008-P, Figure 4.3-60, compares the pressurizer level transients for the AOR and RETRAN-3D calculations. Using the RETRAN-3D calculation as representative, the pressurizer level generally increases from 0 to 68 seconds (denoted here as the insurge period) and decreases from 68 to 79 seconds (denoted here as the outsurge period).
Figure RAI-21-1 overlays the RETRAN-3D pressurizer pressure transient with the RETRAN-3D pressurizer saturation and liquid temperature transients. According to the temperature curves, the pressurizer liquid subcooling: (a) is initially zero; (b) increases gradually during most of the insurge period to about 35°F, due primarily to the inflow of colder water from the pressurizer surge line; (c) decreases rapidly late in the insurge period and early in the outsurge period, due primarily to the effect of decreasing pressure on saturation temperature; and (d) returns to zero for the remainder of the simulation. The transition from (c) to (d) occurs at 72 seconds and is followed by flashing in the pressurizer liquid region and a decrease in the depressurization rate.
[
RA-16-0034 Page 54 of 65
]a,c RA-16-0034 Page 55of65 Figure RAl-21 RNP URBWAP Event-Pressurizer Pressure and Temperature 2400 2200
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RA-16-0034 Page 56 of 65 NRC RAI 22 Section 5 of DPC-NE-3008-P presents VIPRE-01 models that have been expanded relative to the previously-approved models for HNP and RNP discussed in DPC-NE-2005.
- a. Duke Energy stated that these models are available as an option for licensing applications along with the continued use of generic models that use fewer subchannels. Since the NRCs safety evaluation requires VIPRE-01 users to describe the intended use of the code, what licensing applications are these expanded VIPRE-01 models intended to evaluate? Will they be submitted for future NRC review and approval?
- b. The pin power distribution to be used in the expanded VIPRE-01 models is described in Section 5.3, where Duke Energy states that the cycle-specific reactor physics calculations of pin power distributions will be used with appropriate uncertainty factors applied. The approach taken is said to be similar to the approach described for Oconee in DPC-NE-3000, Appendix E.
VIPRE-01 SE condition 3 requires users of the code to submit documentation describing how they intend to use VIPRE-01 and providing justification for their specific modeling assumptions. Given that the Oconee expanded VIPRE-01 model is different from those for HNP and RNP, and that the approach to determining power distribution is only similar to that used at Oconee, specify in additional detail how the power distribution will be determined for the HNP and RNP expanded VIPRE-01 models.
- c. Section 5.4 provides an evaluation of the VIPRE-01 safety evaluation conditions and limitations. VIPRE-01 SE condition 3 requires users of the code to submit documentation describing how they intend to use VIPRE-01 and providing justification for their specific modeling assumptions. In dispositioning this condition on the use of VIPRE-01, Duke Energy refers to the statistical core design methodology described in Revision 5 of DPC-NE-2005. Are the HNP and RNP expanded VIPRE-01 models intended to be used with the statistical core design methodology? Are all the models,
RA-16-0034 Page 57 of 65 correlations, input values of plant specific data, and uncertainties to be applied as described in DPC-NE-2005, Rev. 5? Discuss any differences.
- d. Regulatory Guide 1.203, Section 3.6, states that methodology documentation must discuss nodalization rationale. Other Duke Energy applications of VIPRE-01 (for example, those described in DPC-NE-3000-P-A and DPC-NE-2005-P-A) provided axial and radial nodalization sensitivity studies. Did Duke Energy perform similar studies for the expanded VIPRE-01 models for HNP and RNP? If not, why not? If so, discuss the results of these studies.
- e. VIPRE-01 includes coefficients for crossflow momentum transfer and turbulent mixing between adjacent subchannels. Selection of these coefficients is important, particularly when modeling adjacent fuel assemblies from different fuel vendors, as appears to be an intended application of the expanded VIPRE-01 models for HNP and RNP.Section III.3 of SRP 4.4 instructs the NRC staff to determine that the values of void, pressure drop, and heat transfer correlations used to estimate fluid conditions are within the ranges of applicability specified by their authors or in previous staff reviews. However, these coefficients were not discussed in DPC-NE-3008-P. Justify how Duke Energy determined the crossflow and turbulent mixing coefficients. Explain how the choice of these coefficients is validated, particularly for mixed-core applications.
- f. The importance of model assessment is discussed throughout SRP 15.0.2, Regulatory Guide 1.203, and SRP 4.4. However, Duke Energy provided no benchmarking and validation of the expanded VIPRE-01 models for HNP and RNP. How will these models be validated, particularly when used in mixed-core applications where limited data may be available?
RA-16-0034 Page 58 of 65 Duke Energy RAI 22 Response Response to RAI-22a The expanded models for Robinson Nuclear Plant (RNP) and Harris Nuclear Plant (HNP) provide two main features. The first feature enables the detailed geometric representation of the [
]a,c. In addition, the [
]a,c can be modeled. This provides modeling capabilities that are not possible with the [
]a,c models described in Reference RAI-22-3. The second feature is the option to use the SIMULATE-3 physics code predicted cycle-specific core and pin power distributions. Both of these features typically result in DNB margin gains. The expanded models can also be used to assess advances in fuel assembly design such as radial zoned enrichment and different burnable poison rod loading patterns and issues such as fuel assembly bow, which benefit by modeling the actual core and pin power distributions rather than use of generic power distribution inputs to determine DNB impact.
One of the main applications of the expanded model is to perform cycle-specific DNB analysis.
Duke uses the Maximum Allowable Peaking (MAP) methodology described in References RAI-22-1 and RAI-22-2 to perform the reload core DNB analyses to ensure the DNBR limits are not exceeded. The MAP limits are a family of curves, typically represented as maximum allowable peak versus axial location of peak, with axial peaking factor as a parameter. These curves are a locus of points for which the minimum DNBR is equal to the analysis statistical DNB limit (SDL) plus retained margin. The MAP limits are calculated using the VIPRE-01 [
]a,c models and applying the conservative reference radial pin power distributions shown in Reference RAI-22-3 (Appendix H, Figure H-2 for RNP and Appendix I, Figure I-2 for HNP).
The reference pin power distributions contain a cluster of pins near the targeted limiting DNB sub-channel that are at a higher power relative to the rest of the pins in the fuel assembly. This results in a flatter pin power distribution near the limiting DNB sub-channel, which results in less mixing by neighboring cooler sub-channels. Both the flatter power distribution and reduced mixing result in a more conservative DNBR calculation.
RA-16-0034 Page 59 of 65 For each reload core design, accident specific MAP limits are compared to predicted peaking results from Chapter 15 non-LOCA transients to confirm acceptable DNB performance. If negative peaking margin (predicted pin peaking greater than the appropriate MAP limit) is determined, the MDNBR may be calculated using the expanded models. This analyses use the three dimensional power distribution from SIMULATE-3. The case specific SIMULATE-3 radial pin power distribution and axial power distribution are input directly into the expanded model and the MDNBR calculated. Since the critical heat flux correlation and the SDL are approved in Reference RAI-22-3 and the expanded models will be approved through this LAR, no additional submittals to the NRC are required for this application.
Another intended application of the expanded model is for calculating the DNB mixed core penalty due to a fuel design transition. The methodology for determining the mixed core penalty using the expanded model will be separately submitted for NRC review and approval prior to implementation.
Response to RAI-22b The expanded models feature more sub-channels than the previously approved VIPRE-01 models for HNP and RNP and thus facilitates modeling cycle-specific fuel assembly and pin power distributions. As described in response to Question 22.a, the MAP methodology uses conservative reference pin power distributions shown in Reference RAI-22-3. If negative peaking margin occurs, a statepoint specific MDNBR analysis is performed using the SIMULATE-3 physics code predicted cycle-specific radial pin power distribution and axial power distribution with appropriate uncertainties applied. The SIMULATE-3 radial pin power distribution for the expanded model is constructed using the same method described in Reference RAI-22-4, Appendix E, for ONS.
First, for each fuel pin, the difference between the maximum pin peaking factor in that assembly is calculated. Second, a multiplier is selected and applied to this difference to produce a smaller adjusted delta-power value for each pin. Lastly, the adjusted delta-power values are subtracted from the maximum pin peaking factor to create the final revised peaking value for each specific fuel pin. The result of this final adjustment is flatter radial pin power distribution which preserves the peak pin value for the assembly. This process results in a more conservative pin power distribution for input to VIPRE-01 relative to the unadjusted radial power distribution case.
RA-16-0034 Page 60 of 65 Response to RAI-22c The expanded models will not be used to generate the statistical DNB limits (SDL) using the statistical core design (SCD) methodology. The expanded models will be used in analyses with the calculated SDL from Reference RAI-22-3 for HNP and RNP. All the VIPRE-01 input parameters not related to [
]a,c for HNP and RNP will be applied as described in DPC-NE-2005, Revision 5. No differences in these inputs are needed for application to, or use of, the expanded models. The expanded models add increased radial definition of local hydraulic effects of additional sub-channels and the radial pin by pin peaking distribution to more realistically model the peaking and thermal-hydraulic conditions within the hot fuel assembly. The expanded models geometric change do not impact the statistical DNB behavior of the fuel design.
Response to RAI-22d Radial Nodalization:
The expanded model represents [
]a,c The [
]a,c as depicted in Figure 5.1-1 (Reference RAI-22-5) for the HNP W17-HTP fuel assembly and Figure 5.2-1 (Reference RAI-22-5) for the RNP W15-HTP fuel assembly. Duke sensitivity studies previously performed in Reference RAI-22-1, RAI-22-2, and RAI-22-4 have shown that the hot sub-channel flow conditions are not sensitive to how the rest of the hot bundle and remaining core are modeled. The previous Duke studies independently confirm the radial nodalization sensitivity study documented in the VIPRE-01 code manual (Reference RAI-22-6) which concludes that at least one full row of sub-channels completely surrounding the hot sub-channel is necessary to adequately resolve the details of the flow field in the vicinity of the hot sub-channel. If that detail is provided by the model, the hot sub-channel flow conditions are not sensitive to how the rest of the hot bundle and remaining core are modeled. Because of the extensive sub-channel detail of the expanded model, a radial nodalization sensitivity analysis was not performed.
RA-16-0034 Page 61 of 65 Axial Nodalization:
The VIPRE-01 predictions are sensitive to axial nodalization in that enough nodes must be provided to resolve the detail in the axial power profile and the flow field. Axial nodalization sensitivity studies have been performed using both the McGuire/Catawba and Oconee [
]a,c models documented in Reference RAI-22-4. The McGuire/Catawba sensitivity study used uniform node sizes of [
]a,c inches per active fuel node. The Oconee sensitivity study used uniform node sizes of [
]a,c inches per active fuel node. Results from these studies concluded that once sufficient accuracy is obtained, the VIPRE-01 prediction becomes insensitive to further axial node refinement. The Duke study results are consistent with Reference RAI-22-6 axial noding sensitivity study which modeled four cases using uniform node sizes of 4 inches, 6 inches, 8 inches and 12 inches and one case using non-uniform axial node sizes. In the non-uniform case, the node size was 8 inches for the first 48 inches, 6 inches from 48 inches to 60 inches, and 4 inches for the remainder of the axial length (where the MDNBR was expected to occur). The uniform node size sensitivity study showed no significant differences in MDNBR results amongst the cases. The non-uniform case result was essentially equivalent to the results of the uniform case with 4-inch nodes. The previous Duke study results confirmed the findings in Reference RAI-22-6.
The expanded RNP and HNP models utilizes identical axial nodalization scheme as their respective [
]a,c models, approved in Reference RAI-22-3 (Appendices H and I, respectively). The RNP and HNP [
]a,c models, are based on the Oconee [
]a,c model and use non-uniform node sizes with the largest node size no greater than
[
]a,c inches, respectively. These noding sizes are finer than the node sizes utilized in the two sensitivity studies quoted. Therefore, it was concluded that the VIPRE-01 prediction is insensitive to further axial node refinement for HNP and RNP. No additional sensitivity analysis was performed for the expanded models.
Response to RAI-22e Turbulent mixing is a sub-channel phenomenon and is conservatively assumed not applicable to fuel assembly boundaries or to lumped channels. The VIPRE-01 energy and momentum equations contain terms that describe exchange of energy and momentum between adjacent
RA-16-0034 Page 62 of 65 sub-channels due to turbulent mixing. The following are the two key elements of the turbulent mixing model:
Turbulent Momentum Factor (FTM):
The FTM specifies how efficiently the turbulent crossflow mixes momentum. It can be specified on a scale from 0.0 to 1.0, where 0.0 specifies that the turbulent crossflow mixes enthalpy only and not momentum and 1.0 specifies that it mixes momentum with the same strength as it mixes enthalpy (Reference RAI-22-6). The expanded models, consistent with the respective
[
]a,c models, [
]a,c This parameter is not fuel assembly dependent and therefore requires no special treatment for mixed core applications.
Turbulent Mixing Coefficient (ABETA):
Turbulent crossflow (w') is a function of the sub-channel gap width and the average of the flow in the adjacent sub-channels (Reference RAI-22-6). The associated turbulent mixing coefficient, ABETA, is fuel assembly type dependent and is provided by the fuel vendor. The turbulent mixing coefficient is applied only to sub-channels within the same fuel assembly and is set to zero in the sub-channels between assembly boundaries in the expanded models. The expanded models are designed to accommodate different turbulent mixing coefficients for individual fuel assemblies if a different design is located next to the hot fuel assembly, as may occur in a mixed core application. The appropriate fuel vendor recommended value would be used for each specific fuel design.
The use of the turbulent mixing coefficient in the expanded models for the mixed core application due to a fuel design transition will be separately submitted for NRC review and approval prior to implementation.
Response to RAI-22f Due to lack of appropriate measured data to benchmark against and the proprietary nature of the other analyses that model detailed fuel assembly sub-channels, Duke validated the expanded models against their respective RNP and HNP VIPRE-01 [
]a,c models RA-16-0034 Page 63of65 approved in DPC-NE-2005-PA (Appendix H for RNP and Appendix I for HNP Reference RAl-22-3).
The validation used the same statepoint conditions and radial and axial power distributions for both models of each plant. All other code options, correlations and parameters applied were identical between the [
]a,c and the expanded models. This evaluation to assess model performance used a broad range of conditions via selected UFSAR/FSAR Chapter 15 non-LOCA event statepoint conditions and cycle-specific axial and radial pin power distributions. The results are summarized in the tables below:
TABLE RAl-22-1 ROBINSON NUCLEAR PLANT Transient Discription
[
1 MDNBR
[
Increase in Steam Flow Loss of External Load Loss of Flow Unc RCCA Bank w/d (BOC HFP)
Unc RCCA Bank w/d (EOC HFP)
Unc RCCA Bank w/d (BOC 600'6 FP)
I Unc RCCA Bank w/d (EOC 60% FP)
Misaligned RCCA or RCCA Bank Dropped Rod Inadvertent Opening of PORV AVERAGE DIFFERENCE=
1 Difference a,c a,c
[
1 a,c RA-16-0034 Page 64of65 Transient Discription Increase in Steam Flow Loss of External Load Loss of Flow Unc RCCA Bank w/d (BOC HFP)
Unc RCCA Bank w/d (EOC HFP)
TABLE RAl-22-2 HARRIS NUCLEAR PLANT MDNBR
[
1
[
Unc RCCA Bank w/d (BOC 60"/o FP)
Unc RCCA Bank w/d (EOC 60"/o FP)
Misaligned RCCA or RCCA Bank Dropped Rod Inadvertent Opening of PORV AVERAGE DIFFERENCE=
1 Difference a,c a,c
~
[
1 a,c The results show comparable and consistent performance of the expanded model to the (
]a,c model for both the RNP and HNP. While not the intended purpose, this comparison does illustrate the conservative nature of the [
model.
]8*c model relative to the expanded The evaluation of the expanded models for the mixed core applications due to a fuel design transition will be separately submitted for NRC review and approval prior to implementation.
RA-16-0034 Page 65 of 65 References RAI-22-1 DPC-NE-2003-PA, Revision 3, Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01, April 2012.
RAI-22-2 DPC-NE-2004-PA, Revision 2a, McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01, December 2008.
RAI-22-3 DPC-NE-2005-P, Revision 5, Duke Energy Thermal-Hydraulic Statistical Core Design Methodology, (as approved by Safety Evaluation dated March 8, 2016).
RAI-22-4 DPC-NE-3000-PA, Revision 5a, Oconee, McGuire and Catawba Nuclear Station Thermal-Hydraulic Transient Analysis Methodology, October 2012.
RAI-22-5 DPC-NE-3008-P, Revision 0, Shearon Harris and H. B. Robinson Nuclear Plants Thermal-Hydraulic Models for Transient Analysis, November 2015.
RAI-22-6. EPRI NP-2511-CCM, Revision 3, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores", TAC No. M79498, October 1993.