PROPRIETARY INFORMATION – WITHHOLD UNDER …JOSEPH DONAHUE Vice President Nuclear Engineering 526...
Transcript of PROPRIETARY INFORMATION – WITHHOLD UNDER …JOSEPH DONAHUE Vice President Nuclear Engineering 526...
JOSEPH DONAHUE Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202
980-373-1758 [email protected]
PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED
PROPRIETARY INFORMATION – WITHHOLD UNDER 10 CFR 2.390
UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED Serial: RA-17-0043 10 CFR 50.90 October 9, 2017 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-3009, REVISION 0
REFERENCES: 1. Duke Energy letter, Supplemental Information for License Amendment Request
Regarding Methodology Report DPC-NE-3008-P, dated October 3, 2016 (ADAMS Accession No. ML16278A080)
2. NRC letter, Duke Energy Progress, LLC, For Shearon Harris Nuclear Power Plant, Unit 1, and H. B. Robinson Steam Electric Plant, Unit No. 2 - Request For Additional Information Regarding Application to Adopt DPC-NE-3008-P, Revision 0, "Thermal-Hydraulic Models for Transient Analysis," and DPC-NE -3009-P, Revision 0, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology" (CAC Nos. MF8439 and MF8440), dated September 8, 2017 (ADAMS Accession No. ML17226A264)
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 supplemental 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). In part, Duke Energy requested NRC review and approval of DPC-NE-3009-P, Revision 0, “FSAR / UFSAR Chapter 15 Transient Analysis Methodology,” and adoption of the methodology into the TS for HNP and RNP. In Reference 2, the NRC requested additional information (RAI) regarding DPC-NE-3009.
Attachment 3 provides Duke Energy’s response to a portion of the Reference 2 RAI’s. The specific RAI’s included are: RAI-3, 5, 6, 8-13, 15-17, 21, 22, 24, 25, 28, 29, 31-35, 37-39, 42,
PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED
U.S. Nuclear Regulatory Commission RA-17-0043 Page2
45-47, 49, and 53-57. Responses to the remaining RAl's will be provided at a later date. Attachment 3 contains information that is proprietary to Duke Energy. In accordance with 1 O CFR 2.390, Duke Energy requests that Attachment 3 be withheld from public disclosure. An affidavit is included (Attachment 1) attesting to the proprietary nature of Attachment 3. A nonproprietary version of Attachment 3 is included in Attachment 2.
This submittal contains no new regulatory commitments. 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 October 9, 2017.
Sincerely,
~v:i~~ Joseph Donahue Vice President - Nuclear Engineering
JBD
Attachments: 1. Affidavit of Joseph Donahue 2. Response to NRC Request for Additional Information (Redacted) 3. Response to NRC Request for Additional Information (Proprietary)
cc: (all with Attachments unless otherwise noted)
C. Haney, Regional Administrator USNRC Region II J. Zeiler, USNRC Senior Resident Inspector - HNP J. Rotton, USN RC Senior Resident Inspector - RNP M. C. Barillas, NRR Project Manager - HNP D. J. Galvin, NRR Project Manager - RNP W. L. Cox, Ill, Section Chief, NC DHSR (Without Attachment 3) S. E. Jenkins, Manager, Radioactive and Infectious Waste Management Section (SC)
(Without Attachment 3) A. Wilson, Attorney General (SC) (Without Attachment 3) A. Gantt, Chief, Bureau of Radiological Health (SC) (Without Attachment 3)
PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED
Attachment 1 RA-17-0043
Attachment 1
Affidavit of Joseph Donahue
Attachment 1 RA-17-0043 Page 1 of 3
AFFIDAVIT of Joseph Donahue
1. I am Vice President of Nuclear Engineering, 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 Energy’s 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 Attachment 3 to Duke Energy RAI response letter RA-17-0043 regarding application to revise technical specifications for report DPC-NE-3009-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 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 reveals cost or price information,
production capacities, budget levels or commercial strategies of Duke Energy or its customers or suppliers.
Attachment 1 RA-17-0043 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 information’s 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
Attachment 3 to Duke Energy RAI response letter RA-17-0043 regarding application to revise technical specifications for report DPC-NE-3009-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.
Attachment 1 RA-17-0043 Page 3 of 3
Joseph Donahue 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 October 9, 2017.
~~~~ JQSePhOnahue
Attachment 2 RA-17-0043
Attachment 2
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.
Attachment 2 RA-17-0043 Page 1 of 70
NRC RAI 3
Duke Energy proposed in DPC-NE-3008 that “[l]icensing 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.” While Section 5.0 of DPC-NE-3009 proposed
potential applications for this model, these uses were not described in sufficient detail for the
NRC staff to review them. The issue is not that the model is suspect – indeed, being able to add
heat to a fluid volume without using a heat conductor is a standard feature of thermal-hydraulic
systems analysis codes. However, the future uses were not specified in sufficient detail for the
NRC staff to determine whether or not they are acceptable. Please provide additional detail for
the NRC staff to complete the review. (SRP 15.0.2)
Duke Energy RAI 3 Response
Non-conducting heat exchangers may be used to model ambient losses from the primary
system (including the pressurizer) in certain long-term transients such as Loss of Non-
Emergency AC Power to the Station Auxiliaries, Loss of Normal Feedwater Flow and Feedwater
System Pipe Break. The HNP/RNP RETRAN-3D base models assume that all heat conductors
in contact with the containment atmosphere have a perfectly insulated boundary at that side
such that no heat transfer occurs from the metal surfaces of the primary system to the
containment atmosphere. To account for the fact that some energy is transferred from the metal
surfaces through the insulation and into the containment atmosphere, non-conducting heat
exchangers may be attached to the primary system. [
]a,c
Attachment 2 RA-17-0043 Page 2 of 70
Non-conducting heat exchangers may also be used when a more detailed feedwater system
model is desired for transients such as Steam System Piping Failure. Additional volumes and
junctions are used to model the feedwater piping, with the feedwater heaters simulated using
non-conducting heat exchangers. The heat addition rates from each feedwater heater are
determined using plant design data and are adjusted to conservatively bound expected system
alignments of the condensate and main feedwater systems, consistent with the initial condition
of the event.
When the simplified steam generator secondary model is used for transients that are initiated
from hot zero power, it may be desirable to use non-conducting heat exchangers to remove the
heat generated by the operating reactor coolant pumps during the steady-state initialization.
When this method is employed, there is no flow modeled in the secondary side during the
steady-state initialization. The non-conducting heat exchangers are added to the [
]a,c This method simplifies the
steady-state initialization process and results in initial conditions equivalent to those obtained
using feedwater addition for RCS heat removal during the steady-state initialization.
Attachment 2 RA-17-0043 Page 3 of 70
NRC RAI 5
Duke Energy proposed in Section 5.0 to use the RETRAN-3D decay heat model. In specifying
how this model will be used, Duke Energy did not specify which transients would use low,
nominal, or high decay heat assumptions. Please clarify the decay heat assumptions for the
transients considered in DPC-NE-3009. (SRP 15.0.2)
Duke Energy RAI 5 Response
Most of the non-LOCA transients covered by DPC-NE-3009 are short in duration and reach the
limiting conditions prior to or near the time of reactor trip. For these transients, the decay heat
assumption is not important and is assumed to be nominal (1979 ANS Standard with a multiplier
of 1.0).
There are only five long-term transients in DPC-NE-3009 for which the decay heat assumption
is important. Decay heat for these transients will be treated as follows:
Transient Decay Heat Assumption
Steam System Piping Failure Low decay heat is conservative in order to maximize the
overcooling of the primary system.
Loss of Non-Emergency AC
Power to the Station Auxiliaries
High decay heat is conservative in order to maximize the
post-trip heatup of the primary system.
Loss of Normal Feedwater Flow
Feedwater System Pipe Break
Steam Generator Tube Rupture For the overfill analysis, sensitivity studies will be performed
to determine if high or low decay heat is limiting. For the
thermal-hydraulic input analysis, high decay heat is
conservative in order to maximize the post-trip heatup of the
primary system.
Attachment 2 RA-17-0043 Page 4 of 70
NRC RAI 6
Section 3.3, “VIPRE-01,” proposes the use of the “Modified Barnett” correlation for critical heat
flux evaluations of off-rated conditions in VIPRE-01. The cited basis for approval of this
correlation is Section 15.3.1.2.5 of DPC-NE-3005, “UFSAR Chapter 15 Transient Analysis
Methodology,” (Reference 13). The NRC staff found no mention of the Modified Barnett
correlation in the version of the referenced topical report it reviewed (Revision 2). However,
Section 8.0, “References,” lists DPC-NE-3005, Revision 5, which has not been submitted to the
NRC. Please provide additional justification for the use of the Modified Barnett correlation or
provide evidence that it has been reviewed and approved by the NRC staff for similar
applications. (SRP 4.4)
NRC RAI 6 Reference
13. DPC-NE-3005-PA, “UFSAR Chapter 15 Transient Analysis Methodology,” Revision 2,
Duke Power Co., May 2005 (ADAMS Accession No. ML051740176 (proprietary) and
ML051740156 (non-proprietary)).
Duke Energy RAI 6 Response
DPC-NE-2015-PA, Revision 0 (Reference RAI-6-1), describes the methodology for transitioning
to the Mark-B-HTP fuel assembly design at the Oconee Nuclear Station. Section 7.3 of
Reference RAI-6-1 describes a new Appendix D for DPC-NE-3000-PA, which describes
thermal-hydraulic models and benchmark analyses for the Catawba, McGuire and Oconee
Nuclear Stations. One of the changes described in the new Appendix D was to incorporate the
Modified Barnett critical heat flux correlation into the Main Steam Line Break analysis
methodology. Section 3.4 of the Safety Evaluation Report (SER) for DPC-NE-2015-PA,
Revision 0 (Reference RAI-6-2), addresses the new Appendix D and acknowledges the use of
the Modified Barnett correlation for the Main Steam Line Break analysis. According to the SER,
“The NRC staff concludes that Appendix D is acceptable to incorporate into approved
methodology report DPC-NE-3000-PA.”
Attachment 2 RA-17-0043 Page 5 of 70
Duke Energy RAI 6 Response References
RAI-6-1. “Mark-B-HTP Fuel Transition Methodology,” DPC-NE-2015-PA, Revision 0, October
2008.
RAI-6-2. Olshan, L. N., to Baxter, D., “Oconee Nuclear Station, Units 1, 2, and 3, Issuance of
Amendments Regarding Use of AREVA NP Mark-B-HTP Fuel (TAC Nos. MD7050,
MD7051, and MD7052),” October 29, 2008. (ADAMS Accession No. ML082800408)
Attachment 2 RA-17-0043 Page 6 of 70
NRC RAI 8
In Section 4.0, “Safety Analysis Physics Parameters,” Duke Energy proposed that the bounding
nature of the safety analysis can be determined by comparing key physics parameter values
from previous, “similar” core reload designs to the safety analysis values. How does the
proposed method determine that one core reload design is “similar” to another core reload
design? Additionally, how does such an approach demonstrate the amount of margin available
between the proposed core design and the safety analysis? (SRP 15.0.2)
Duke Energy RAI 8 Response
A core design is considered similar if its [
]a,c
Attachment 2 RA-17-0043 Page 7 of 70
An example where this option could be used is for a core redesign resulting from the
identification of a damaged fuel assembly during refueling. In this scenario, the energy
requirements (in terms of the number and enrichment of fresh fuel assemblies) and burnable
absorber requirements are already established. An exchange of the damaged fuel assembly
with one of similar reactivity is typically performed with minimal or no reinsert fuel shuffling
required. In this instance, the similarity of the loading patterns would be determined based on
the comparisons described above, and the [
]a,c
Attachment 2 RA-17-0043 Page 8 of 70
NRC RAI 9
In Section 4.1, “Generic Parameters,” Duke Energy proposed the use of a “conservative
relationship between rod position (percent inserted) and normalized reactivity worth.” Please
identify how this relationship is determined. (SRP 4.3)
Duke Energy RAI 9 Response
The position-dependent negative reactivity insertion 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. Both the inserted worth and
rod position domains are normalized to produce the normalized worth versus normalized rod
position curve. Conservatism introduced by the bottom-peaked power distribution is dependent
upon the magnitude of the axial flux difference operating limits. An example normalized trip
reactivity worth versus normalized rod position curve for Harris is shown in Figure RAI-9-1. The
trip reactivity shape curve is multiplied by the minimum trip worth to obtain trip reactivity versus
rod position. A cycle-specific reload check is performed to verify the acceptability of the
minimum trip reactivity versus rod position curve assumed in the safety analysis.
Attachment 2 RA-17-0043 Page 9of70
Figure RAl-9-1 - Example Trip Reactivity Shape (Normalized Worth vs. Normalized Rod
Position)
a,c
Attachment 2 RA-17-0043 Page 10 of 70
NRC RAI 10
In Section 4.1, Duke Energy proposed that “perturbed power distributions allowed by the AFD
[Axial Flux Difference] and rod insertion limits are considered” for transients in which the initial
power distribution significantly impacts the course of the event. It is unclear to the NRC staff
what these “perturbed power distributions” are and how they are different from the “core power
distributions permitted by AFD and rod insertion limits” discussed earlier in the same paragraph.
Please clarify. (SRP 15.0.2, values of core physics parameters in Chapter 15 sections)
Duke Energy RAI 10 Response
The perturbed power distributions referenced in the last sentence of Section 4.1, Sub-Section
“Initial Power Distribution,” are the same as those permitted by the AFD and rod insertion limits
used to ensure the acceptability of F∆H and FQ limits. The intent of the last sentence is to convey
that for transients such as an uncontrolled RCCA bank withdrawal, the initial core power
distribution is important not only from a peaking factor perspective, but also because it
significantly impacts the course of the event by influencing the amount of reactivity available for
withdrawal and the rate of reactivity insertion.
Attachment 2 RA-17-0043 Page 11 of 70
NRC RAI 11
In Section 4.1, Duke Energy stated that the prompt neutron lifetime is not a key parameter and
typical time-in-life values would be used for it. However, in the NRC-approved methodology
described in DPC-NE-3001, “Multidimensional Reactor Transients and Safety Analysis Physics
Parameters Methodology” (Reference 14), this parameter is biased such that the ratio of beta-
effective to the prompt neutron lifetime is minimized, which increases the neutron power spike in
the event of prompt criticality. Why is this not necessary in DPC-NE-3009? (SRP 15.0.2, values
of core physics parameters in Chapter 15 sections)
NRC RAI 11 Reference
14. DPC-NE-3001-PA, “Multidimensional Reactor Transients and Safety Analysis Physics
Parameters Methodology,” Revision 0a, Duke Energy Carolinas, LLC, May 2009
(ADAMS Accession No. ML16102A168 (proprietary) and ML16102A158 (non-
proprietary)).
Duke Energy RAI 11 Response
Section 5.0 of DPC-NE-3001-PA, Revision 0a (Reference RAI-11-1), describes the method for
analyzing the event denoted in DPC-NE-3009 as Steam System Piping Failure for the Catawba
and McGuire Nuclear Stations. According to Sub-Section 5.3.2.6 of Reference RAI-11-1, “the
prompt neutron lifetime value is chosen to minimize the ratio of beta-effective to the prompt
neutron lifetime. This ... increases the neutron power spike when prompt criticality is achieved.”
This approach represents a conservative exception to the general treatment of prompt neutron
lifetime as “typical”, as described in Section 2.2 of Reference RAI-11-1. Justification for the
general treatment was provided in the response to RAI 27 of Reference RAI-11-1 for three
events initiated by uncontrolled RCCA withdrawals. The Safety Evaluation Report
acknowledged the response and concluded that “the determination and application of the key
safety parameters in the DPC reload methodology are acceptable.” These references support
the general treatment of prompt neutron lifetime as “typical”, as proposed in Section 4.0 of DPC-
NE-3009.
Attachment 2 RA-17-0043 Page 12 of 70
In order to resolve RAI 11 on DPC-NE-3009, [
]a,c
Duke Energy RAI 11 Response Reference
RAI-11-1. “Multidimensional Reactor Transients and Safety Analysis Physics Parameters
Methodology,” DPC-NE-3001-PA, Revision 0a, May 2009.
Attachment 2 RA-17-0043 Page 13 of 70
NRC RAI 12
The methods used to determine several key physics parameters in Section 4.2, “Control Rod
Worth Calculations” (maximum differential rod worth at subcritical, maximum differential rod
worth at power, ejected rod worth) state that adverse power distributions are considered. Please
specify the range of adverse power distributions considered in the evaluation of these
parameters and how they are generated. (SRP 15.0.2, values of core physics parameters in
Chapter 15 sections)
Duke Energy RAI 12 Response
The xenon distributions used to create adverse power distributions for use in control rod worth
calculations are developed by either performing a series of xenon transients or by using a
reference xenon distribution skewed using [
]a,c
Attachment 2 RA-17-0043 Page 14 of 70
Control rod worth calculations are performed [
]a,c This process is repeated for each time
in life where the analysis is performed. The Technical Specification AFD and rod insertion limits
that are assumed in the analysis are specified in the Core Operating Limits Report.
Attachment 2 RA-17-0043 Page 15 of 70
NRC RAI 13
The specification of ejected rod worth in Section 4.2 indicates that only certain limiting locations
are considered in the analysis. The previously-approved DPC-NE-3001 methodology states that
“all possible rods” are analyzed. Without considering all locations, a potentially limiting rod might
be missed. Clarify how limiting locations are determined so that a potentially limiting rod is not
missed. (SRP 15.0.2, values of core physics parameters in Chapter 15 sections)
Duke Energy RAI 13 Response
The change in the wording in the DPC-NE-3009 report relative to the DPC-NE-3001 report was
made to exclude Control Bank locations that are non-limiting. For example, the Control Bank B
insertion for Robinson at the HZP initial condition is 11 steps. This rod insertion is insufficient to
produce an ejected rod worth that would challenge accident analysis assumptions or produce a
limiting power excursion, and therefore would be excluded from analysis.
In the initial application of the method, all Control Bank D and C rods are evaluated, crediting
core symmetry, to ensure the maximum ejected rod worths assumed in the SIMULATE-3K
accident analysis remain bounding. However, as experience is gained, some Control Bank D
and C core locations may be excluded from future cycle-specific evaluations if they are
demonstrated to be non-limiting. The current control rod insertion limits for Harris and Robinson
are presented in Tables RAI-13-1 and RAI-13-2 for information. These limits are specified in
each plant’s Core Operating Limits Report.
Attachment 2 RA-17-0043 Page 16 of 70
Table RAI-13-1 – Harris Power-Dependent Rod Insertion Limits
Power Level (%) Control Bank D Control Bank C
100 186 226
52.7 98 226
20 37 165
0 0 128
Table RAI-13-2 – Robinson Power-Dependent Rod Insertion Limits
Power Level (%) Control Bank D Control Bank C Control Bank B
100 165 226 226
67.5 98 226 226
20 0 128 226
5.5 0 98 226
0 0 87 215
Attachment 2 RA-17-0043 Page 17 of 70
NRC RAI 15
In Section 5.0, Duke Energy proposed to use the statistical core design (SCD) methodology
described in the NRC-approved DPC-NE-2005 (Reference 15) methodology for evaluation of
departure from nucleate boiling (DNB). This methodology relies on subchannel analyses
performed at a number of statepoints to determine the statistical core design limit, and as such
this limit is only valid when the analysis is performed at or near these statepoints. If the analysis
statepoint falls outside of the range of the statistical design limit, a non-SCD analysis is
performed. The NRC staff has several questions about both the SCD and non-SCD analyses.
a. Duke Energy proposed that the SCD conditions could be updated to bound statepoints
that would otherwise require non-SCD calculations. The NRC staff is unclear on how this
is consistent with Condition 1 on the approval of DPC-NE-2005, which states that
“[Duke] 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.” How will such justification be provided in the event that the SCD
statepoints are adjusted? (Conformance with limitation and condition on DPC-NE-2005)
b. Duke Energy stated that “the final selection of SCD or non-SCD methodology for a given
transient will depend on the actual analysis results and may differ from the expected
selection as presented here.” However, insufficient information has been provided to the
NRC staff to determine the acceptability of the parameter biasing for transients that are
currently envisioned as SCD transients to be performed as non-SCD analyses. This is
because the tables in Section 5 of the methodology report that provide the parameter
biasing simply say “SCD” for key parameters in the DNB ratio (DNBR) evaluation. If the
flexibility to switch between SCD and non-SCD analyses is required, additional
information must be provided to justify the modeling in either scenario for each transient.
(SRP 15.0.2)
Attachment 2 RA-17-0043 Page 18 of 70
NRC RAI 15 Reference
15. Barillas, M., U. S. Nuclear Regulatory Commission, letter to Frisco, J. M., Jr., Duke
Energy Corporation, “Shearon Harris Nuclear Power Plant, Unit 1 and H. B. Robinson
Steam Electric Plant, Unit No. 2 – Issuance of Amendments Revising Technical
Specifications for Methodology Report DPC-NE-2005-P Revision 5, ‘Thermal-Hydraulic
Statistical Core Design Methodology’ (TAC Nos. MF5872 and MF5873),” March 8, 2016
(ADAMS Accession No. ML16049A630).
Duke Energy RAI 15a Response
The Duke Energy SCD methodology in Reference RAI-15-1 directly uses the VIPRE-01
thermal-hydraulic code to do all DNB calculations. The direct use of the code to explicitly
analyze DNBR sensitivity to each unique set of conditions ensures the direct applicability of this
method to varying fuel design and plant parameters.
Section 1.3 of the main body of Reference RAI-15-1 states one of the benefits of direct code
use: “This method can also be used to evaluate a statepoint outside the range of the original
key parameters assumed. If the statepoint statistical DNBR does not exceed the SDL, the
statepoint can apply the licensed limit.” Table 7, SDL Evaluation and Resubmittal Criteria, of
Reference RAI-15-1 lists specific examples of anticipated future conditions analyzed and any re-
submittal requirements before use in license analyses. The third item listed is “New Statepoint”
which requires no further NRC interaction prior to implementation.
Duke Energy RAI 15a Response Reference
RAI-15-1. “Thermal-Hydraulic Statistical Core Design Methodology,” DPC-NE-2005-PA,
Revision 5, March 2016.
Attachment 2 RA-17-0043 Page 19 of 70
Duke Energy RAI 15b Response
In the event that an SCD analysis falls outside the applicable parameter range of the Statistical
Design Limit (SDL), the first recourse is to determine an SDL for that statepoint using the SCD
method as described in the response to RAI 15a. However, in the event that the statepoint
cannot or will not be reanalyzed with the SCD method given in DPC-NE-2005, the transient may
be analyzed using the non-SCD system analysis methodology. The SCD methodology is used
to evaluate the departure from nucleate boiling (DNB) acceptance criteria. (Note that the
remainder of this response will use the term “DNB event” to generically denote a DNB analysis
performed with either the SCD or non-SCD methodology.) DNB events which require analysis
using the non-SCD methodology will typically bias initial core power high, reactor average
temperature high, pressurizer pressure low, reactor coolant system (RCS) flow low, and core
bypass flow high. This parameter biasing will produce a conservative minimum DNBR for most
events. If the DNB event is performed as a non-SCD system analysis, the resultant DNBR from
the core thermal-hydraulic analysis will be compared to the CHF correlation limit to determine
acceptability.
Some DNB events may not use the typical parameter biasing due to the OTΔT reactor trip
function. The OTΔT reactor trip has a dynamic setpoint that changes as a function of reactor
coolant average temperature and pressurizer pressure, which may introduce competing effects.
For example, biasing pressurizer pressure low would typically be conservative for a DNB
analysis as it decreases DNBR. However, relative to using a high initial pressurizer pressure, a
low initial pressurizer pressure would also decrease the OTΔT setpoint, potentially resulting in
an earlier reactor trip and a less limiting statepoint for DNBR. The biasing of pressurizer
pressure and reactor average temperature will therefore be evaluated if a DNB event is
analyzed using the non-SCD methodology and is mitigated by the OTΔT reactor trip function.
Section 5 of DPC-NE-3009 identifies the following DNB events as potentially mitigated by the
OTΔT trip:
Attachment 2 RA-17-0043 Page 20 of 70
● Increase in Feedwater Flow (Section 5.1.1)
● Loss of External Electrical Load (Section 5.2.1)
● Turbine Trip (Section 5.2.2)
● Loss of Normal Feedwater Flow (Section 5.2.4)
● Feedwater System Pipe Break (Section 5.2.5)
● Uncontrolled RCCA Bank Withdrawal at Power (Section 5.4.2)
● Dropped Full-Length RCCA or RCCA Bank (Section 5.4.3)
● Withdrawal of a Single Full-Length RCCA (Section 5.4.4)
● Spectrum of RCCA Ejection Accidents (Section 5.4.8)
● Inadvertent Opening of a Pressurizer Relief or Safety Valve (Section 5.6.1)
● Steam Generator Tube Rupture (Section 5.6.2)
Attachment 2 RA-17-0043 Page 21 of 70
NRC RAI 16
Section 5.0 discussed key parameter selection and biasing, on a generic basis, for all of the
transients. The NRC staff have the following questions about this biasing. (SRP 15.0.2, values
of initial and boundary conditions in Chapter 15 sections)
a. Transients discussed as non-SCD are stated to have uncertainties included in the
biasing of initial and boundary conditions. What uncertainties are included and how are
they represented in the analysis?
b. Does the maximized or minimized reactor coolant system (RCS) flow rate include
instrument and other uncertainties?
c. How is the best estimate of core bypass flow determined?
d. Why is steam generator tube plugging not just assumed to be zero when it is biased
“low”? What is the basis for plugging values assumed that are less than 1%?
Duke Energy RAI 16a Response
Transients analyzed using the non-SCD methodology account for the initial condition
uncertainty in core power, reactor average temperature, pressurizer pressure, total RCS flow,
and core bypass flow. The initial condition uncertainty is calculated using the total loop
uncertainty associated with a given parameter’s automatic control system, where applicable.
The total loop uncertainty is calculated by combining the various component random
uncertainties using the square-root-sum-of-squares (SRSS) method. The component random
uncertainties include primary element, sensor, and rack uncertainty terms, and non-random
uncertainties associated with the system are treated as a bias. If no automatic control system is
associated with a parameter, then the indication uncertainty associated with its Technical
Specification surveillance is used. The treatment of uncertainty for total RCS flow is described in
the response to RAI 16b. A bounding maximum core bypass flow will be used for non-SCD
events.
Attachment 2 RA-17-0043 Page 22 of 70
The parameter uncertainty will be modeled in the RETRAN-3D system analysis by increasing or
decreasing the initial condition by the initial condition uncertainty. Associated control systems for
the parameters are adjusted by the initial condition uncertainty to give a conservative response
and produce bounding consequences for each of the events. No bias is applied to the input to
reactor trips, as the reactor trip analytical limits account for their respective instrument loop
uncertainties.
Duke Energy RAI 16b Response
The maximized and minimized RCS flow rates reflect the uncertainty in the calorimetric
combined with the loop RCS flow measurement uncertainty to create a total RCS flow
measurement uncertainty. The calorimetric is performed after startup to verify that RCS flow is
greater than Technical Specification minimum RCS flow. At HNP and RNP, the surveillance limit
for RCS flow is set to Technical Specification minimum RCS flow plus total RCS flow
measurement uncertainty, ensuring that the minimum measured RCS flow is greater than or
equal to Technical Specification minimum flow. For DNB analyses performed using the SCD
methodology, RCS flow uncertainty is accounted for in the Statistical Design Limit (SDL), so the
RCS flow surveillance limit or minimum measured flow is used as the initial condition. For DNB
analyses performed using the non-SCD methodology, Technical Specification minimum RCS
flow is used as the initial condition. Maximum RCS flow is calculated as maximum design flow
plus total RCS flow uncertainty.
Duke Energy RAI 16c Response
If available, a steady-state thermal-hydraulic analysis of all core bypass flow paths and their
associated geometry and flow resistances will provide the calculated best-estimate core bypass
flow used in the SCD analyses. Otherwise, SCD analyses will assume a conservatively high
core bypass flow that is [
]a,c
Attachment 2 RA-17-0043 Page 23 of 70
Duke Energy RAI 16d Response
The HNP and RNP RETRAN-3D base decks model 0.037% and 0.456% tube plugging,
respectively. These values were calculated using the actual number of plugged and unplugged
steam generator tubes at the time the base models were developed. Tube plugging is expected
to increase with continued plant operation, so the above values for tube plugging are considered
to be conservatively low. As such, these values will be used for transients in which low tube
plugging is conservative.
Attachment 2 RA-17-0043 Page 24 of 70
NRC RAI 17
For the increase in feedwater flow transient discussed in Section 5.1.1, “Increase in Feedwater
Flow,” Duke Energy stated that “where applicable, a coincident step decrease in main feedwater
temperature is also assumed to occur.” It is not clear to the NRC staff how applicability is
determined, or how the step change in feedwater temperature is determined. Please
clarify/justify. (SRP 15.0.2)
Duke Energy RAI 17 Response
The coincident step decrease in main feedwater temperature applies whenever necessary to
bound the transient condition. For HFP cases, the steady-state value typically reflects normal
operation, and a coincident step decrease is applied to bound the transient condition. For HZP
cases, the steady-state value is typically selected to bound the transient condition, and no
coincident step decrease is required. When applicable, the step decrease is calculated as
described in the response to Question 1 in the letter dated 18 August 1995 in DPC-NE-3002-A,
Revision 4b (Reference RAI-17-1). As stated in the response, “The magnitude of the
temperature decrease is conservatively calculated based on maintaining a constant heat
addition rate from the feedwater heaters.”
Duke Energy RAI 17 Response Reference
RAI-17-1. “UFSAR Chapter 15 System Transient Analysis Methodology,” DPC-NE-3002-A,
Revision 4b, September 2010.
Attachment 2 RA-17-0043 Page 25 of 70
NRC RAI 21
Section 5.1.4.1, “RETRAN-3D Models and Options,” proposed the use of the RETRAN-3D
general transport model to track boric acid injected by the emergency core cooling system
(ECCS). Section 4.3, “Reactivity Coefficients,” indicates that differential boron worth is
considered to be a function of reactor coolant temperature and other variables. Given that [[
]], the NRC staff
was unsure of how the differential boron worth was determined for this application. Please
clarify. (SRP 15.0.2, values of core physics parameters in Chapter 15 sections)
Duke Energy RAI 21 Response
A conservatively small differential boron worth was selected for use in the Steam System Piping
Failure accident analysis to bound expected values of this derivative during the course of the
event. SIMULATE-3 differential boron worth calculations were performed at EOC as a function
of [
]a,c
Attachment 2 RA-17-0043 Page 26 of 70
NRC RAI 22
Section 5.1.4.2, “Primary Systems and Components,” proposed to model cold leg accumulators
with the built-in RETRAN-3D accumulator model. However, the NRC staff was unclear on the
initial conditions of the accumulators (including boric acid concentration, initial level, cover gas
pressure) and whether the potential for incursion of noncondensible gases into the RCS is
considered in the steam line break analysis. Please provide additional information on the
modeling and initial/boundary conditions of the accumulators. (SRP 15.0.2, values of initial and
boundary conditions in Chapter 15 sections)
Duke Energy RAI 22 Response
The cold leg accumulators may or may not inject during the time period of interest for a given
(U)FSAR Chapter 15 Steam System Piping Failure transient. Consistent with Section 15.3.1.1.4
of DPC-NE-3005-PA, Revision 2 (Reference RAI-22-1), the initial conditions of the accumulators
(denoted as “core flood tanks” or “CFTs” in Reference RAI-22-1) will be specified using
minimum values of pressure, temperature, inventory and boron concentration. The cold leg
accumulators are not expected to approach an empty condition during the time period of interest
for a (U)FSAR Chapter 15 Steam System Piping Failure transient. Accordingly, the potential for
incursion of non-condensable gases into the RCS is not considered in the analysis.
Duke Energy RAI 22 Response Reference
RAI-22-1. “UFSAR Chapter 15 Transient Analysis Methodology,“ DPC-NE-3005-PA, Revision
2, May 2005.
Attachment 2 RA-17-0043 Page 27 of 70
NRC RAI 24
In the steam line break analysis of Section 5.1.4.3, “Secondary Systems and Components,” in
the methodology report, it is stated that auxiliary feedwater (AFW) is modeled conservatively
and is assumed to be terminated by operator action. (SRP 15.1.5.II, SRP Acceptance Criteria:
“The analyses should take account of the effect that loss of offsite power has … the initiation of
auxiliary feedwater flow, and the effects on the sequence of events for these accidents”; SRP
15.1.5.III: “the availability of the auxiliary feedwater system to supply adequate auxiliary
feedwater flow to the intact steam generators during the accident and the subsequent shutdown
condition is evaluated.”)
a. In the plant analyses of record as presented in the HNP and RNP (U)FSARs, AFW flow
starts at the time of break initiation. Please discuss why it is acceptable to eliminate this
conservatism.
b. No details were provided on the operator action needed to terminate AFW or how (or
even if) it is incorporated into the analysis. Please provide additional details for the NRC
staff’s evaluation.
Duke Energy RAI 24a Response
The current (U)FSAR approach of modeling AFW actuation coincident with break initiation is
judged to be overly conservative. Crediting the time required to generate the AFW actuation
signal is consistent with the typical approach for (U)FSAR Chapter 15 analyses and is judged to
be acceptable. The current (U)FSAR assumption of no time delay following the AFW actuation
signal is conservatively retained.
Attachment 2 RA-17-0043 Page 28 of 70
Duke Energy RAI 24b Response
Operator action is assumed to isolate AFW to the faulted steam generator within 10 minutes
following a main steam line break. This action is consistent with the current (U)FSAR analyses
and is included in the Time Critical Action Program for both plants to ensure that the action can
be accomplished by plant personnel. No new operator action is assumed to isolate AFW.
Attachment 2 RA-17-0043 Page 29 of 70
NRC RAI 25
For the cycle specific evaluation of the steam line break presented in Section 5.1.4.7, “Cycle-
Specific Evaluation,” of the methodology report, Duke Energy stated that [[
]]. Please provide additional details on this analysis.
(SRP 15.0.2)
Duke Energy RAI 25 Response
A description of the power search method used to confirm that the predicted reactivity response
by RETRAN-3D is conservative is described below. The initial step is to determine the critical
eigenvalue prior to performing the power search calculation in SIMULATE-3. Equation RAI-25-1
is used to calculate this eigenvalue:
[ ]a,c (Eq. RAI-25-1)
where:
𝑘𝑘𝑐𝑐𝑖𝑖𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐 = SIMULATE-3 eigenvalue corresponding to the ARI(N-1) condition prior to
transient initiation
∆𝜌𝜌𝑆𝑆𝑆𝑆𝑆𝑆 = Technical Specification minimum shutdown margin (SDM)
∆𝜌𝜌𝑝𝑝𝑝𝑝 = Allowance for the uncertainty in the predicted power defect. This allowance is
calculated by assuming a [
]a,c
Attachment 2 RA-17-0043 Page 30 of 70
The critical eigenvalue from Equation RAI-25-1 [
]a,c
Peaking factors from the core power distribution predicted by SIMULATE-3 are used in
subsequent calculations to confirm the acceptability of DNB and CFM limits.
Attachment 2 RA-17-0043 Page 31 of 70
NRC RAI 28
Section 5.2.3, “Loss of Non-Emergency AC Power to the Station Auxiliaries,” presents the
analysis methodology for the loss of non-emergency alternating current power to the station
auxiliaries transient. NRC staff are asked in SRP 15.2.1-15.2.5 Section III.3 (Reference 17) to
review the effects of single active system or component failures that may affect the course of
the transient. From the licensee's submittal, it is clear that the limiting failure is assumed to
occur in the AFW. However, it is unclear exactly what the limiting failure is or how the AFW is
modeled for the transient. The NRC staff has the following questions to clarify this issue. (SRP
15.2.1-15.2.5)
a. The licensee's submittal states that “Auxiliary feedwater actuates on either low steam
generator level or loss of non-emergency AC power with a conservative delay.” Does
this mean that there is a delay only when AFW actuates on loss of non-emergency AC
power, or is there also one when it actuates on low steam generator level?
b. The licensee's submittal states that “Minimum flow from at least one auxiliary feedwater
pump is assumed.” It is unclear what is implied about the state of the AFW pumps from
this statement, since “at least one auxiliary feedwater pump” having minimum flow says
nothing about the other pumps. Please clarify.
c. What is considered to be the limiting failure within the system? Are a variety of failures
investigated in the analysis to determine what is most limiting?
NRC RAI 28 Reference
17. U. S. Nuclear Regulatory Commission, NUREG-0800, “Standard Review Plan,” Section
15.2.1 - 15.2.5, “Loss of External Load; Turbine Trip; Loss of Condenser Vacuum;
Closure of Main Steam Isolation Valve (BWR [Boiling Water Reactor]); and Steam
Pressure Regulator Failure (Closed),” Revision 2, March 2007 (ADAMS Accession No.
ML070300702).
Attachment 2 RA-17-0043 Page 32 of 70
Duke Energy RAI 28a Response
There is a conservative delay when the Auxiliary Feedwater System (AFW) actuates on either
signal: the loss of non-emergency AC power or the low steam generator level.
Duke Energy RAI 28b and 28c Response
In the Analyses of Record (AORs) discussed below, the Harris and Robinson plants credit the
use of only one AFW pump, which is more conservative than a single failure within the AFW
system. The application of DPC-NE-3009 to the Harris and Robinson plants will evaluate the
conservative treatment of the AFW system (number of steam generators receiving flow and
number of AFW pumps available) to determine whether two AFW pumps may be credited.
For the Harris plant, the worst single failure affecting AFW for the Loss of Non-Emergency AC
Power to the Station Auxiliaries event is the loss of one AFW pump. This single failure leaves
two AFW pumps available. However, the AOR for this event for Harris credits the use of only
one AFW pump.
For the Robinson plant, the Loss of Non-Emergency AC Power to the Station Auxiliaries event
is bounded by other events. For example, the long-term aspects are bounded by the Loss of
Normal Feedwater Flow event with concurrent loss of reactor coolant pumps. The Loss of
Normal Feedwater Flow analysis credits the use of only one AFW pump, which is more
conservative than a single failure for the AFW system.
Attachment 2 RA-17-0043 Page 33 of 70
NRC RAI 29
In the feedwater system pipe break analysis methodology presented in Section 5.2.5,
“Feedwater System Pipe Break,” the short term core cooling and peak primary pressure cases
assume core physics parameters consistent with beginning of cycle (BOC) conditions. The initial
phases of the feedwater line break result in an overcooling event until the faulted steam
generator is depleted. Because of this, it is not clear that BOC core physics parameters would
provide the most limiting response, especially for the short-term core cooling case, where a
more negative moderator temperature coefficient (MTC) could provide for a worse core power
response. Please justify the choice of core physics parameters for the short-term core cooling
and peak primary pressure cases. (SRP 15.2.1-15.2.5.II.3.C, “The core burn-up is selected to
yield the most limiting combination of moderator temperature coefficient, void coefficient,
Doppler coefficient, axial power profile, and radial power distribution.”; SRP 15.2.1-15.2.5.III.5,
“The values of system parameters and initial core and system conditions as input to the model
are reviewed by the organization responsible for reactor systems. Of particular importance are
(A) the values of reactivity coefficients and control rod worths in the applicant's analysis and (B)
the variations of moderator temperature, void, and Doppler coefficients of reactivity with core
life. The reviewer evaluates the applicant’s justification showing that the core burn-up selected
yields the minimum safety margins.”)
Duke Energy RAI 29 Response
Depending on the size of the break and the plant operating conditions at the time of the break,
the feedwater system pipe break could cause either a Reactor Coolant System (RCS) cooldown
(by excessive energy discharge through the break) or heatup (from the loss of the reactor
system heat sink). The methodology described in Section 5.2.5 of DPC-NE-3009 evaluates the
RCS heatup aspects of the Feedwater System Pipe Break transient.
Attachment 2 RA-17-0043 Page 34 of 70
If a Feedwater System Pipe Break analysis were performed to evaluate the RCS cooldown
aspects of the transient, it would use a most-negative moderator temperature coefficient and a
least-negative Doppler temperature coefficient. However, the potential RCS cooldown resulting
from a secondary pipe rupture is bounded by the cooldown in the Steam System Piping Failure
transient. In the Steam System Piping Failure analysis methodology, core physics parameters
are chosen to maximize the return to power in the RETRAN-3D calculations (refer to Section
5.1.4.4 of DPC-NE-3009).
Attachment 2 RA-17-0043 Page 35 of 70
NRC RAI 31
In the partial or complete loss of forced reactor coolant flow analysis methodology described in
Section 5.3.1, “Partial or Complete Loss of Forced Reactor Coolant Flow,” it is mentioned that
the peak pressure evaluation assumes that offsite power is maintained but no such discussion
is provided about the core cooling evaluation. Please provide a discussion of whether or not it is
conservative from a DNBR standpoint to model a loss of offsite power for the loss of flow
transient. (SRP 15.3.1-15.3.2.III.2 (Reference 18): “For new applications, LOOP [loss of offsite
power] should not be considered a single failure; each loss of flow transient should be analyzed
with and without a LOOP in combination with a single active failure.”)
NRC RAI 31 Reference
18. U. S. Nuclear Regulatory Commission, NUREG-0800, “Standard Review Plan,” Section
15.3.1 - 15.3.2, “Loss of Forced Reactor Coolant Flow Including Trip of Pump Motor and
Flow Controller Malfunctions,” Revision 2, March 2007 (ADAMS Accession No.
ML070550010).
Duke Energy RAI 31 Response
The Complete Loss of Forced Reactor Coolant Flow event (denoted as Complete LOF here) is
defined to result from the instantaneous trip of all reactor coolant pumps related to the loss of
power to the pump motors. In the analysis methodology, the availability of offsite power affects
the availability of systems and components in the secondary system (e.g., the Main Feedwater
System). However, the behavior of the secondary system (and availability of offsite power) is
judged to have a negligible effect on the minimum DNBR results. This conclusion is based on
the timing of minimum DNBR compared to the loop transit time.
For the discussion below, the loss of offsite power is assumed to occur coincident with reactor
trip and turbine trip. The assumption that offsite power is lost coincident with turbine trip is
consistent with the treatment of offsite power availability in Section 5.3.2 of DPC-NE-3009.
Attachment 2 RA-17-0043 Page 36 of 70
In the demonstration analysis described in Section 6.3 of DPC-NE-3009, reactor trip occurs at
1.2 seconds, followed by minimum DNBR at 3.2 seconds. A loop transit time greater than 2
seconds (i.e., 3.2 seconds minus 1.2 seconds) in this primary-system-driven event suggests
that secondary system behavior has a negligible effect on the transient results. In the
demonstration analysis described in Section 6.3 of DPC-NE-3009, the loop transit time is
estimated to be greater than 10 seconds with the complete loss of flow.
In contrast to the Complete LOF, a Partial LOF is defined to result from the instantaneous trip of
only one reactor coolant pump. In a Partial LOF analysis, the availability of offsite power affects
the availability of the remaining reactor coolant pumps, as well as the systems and components
in the secondary system. With a loss of offsite power, the remaining reactor coolant pumps
begin to coast down after a short delay following reactor trip and turbine trip. Based on the
amount of flow reduction and the rate of flow decrease in the primary system, modeling a loss of
offsite power (i.e., tripping the remaining reactor coolant pumps) is judged to yield a
conservative minimum DNBR relative to the case with offsite power available. However, the
amount of flow reduction and rate of flow decrease in the Partial LOF analysis are less than in
the Complete LOF analysis. Therefore, the Complete LOF analysis provides a bounding
minimum DNBR compared to the Partial LOF analysis, provided the ANS Condition II
acceptance criteria of the Partial LOF analysis are met.
Attachment 2 RA-17-0043 Page 37 of 70
NRC RAI 32
In the complete loss of flow demonstration analysis presented in Section 6.3, “Complete Loss of
Forced Reactor Coolant Flow (HNP),” of the report, it appears that the reactor coolant pump
coastdown curve is much less conservative than the one assumed in the analysis of record
based on the core flow rate plots provided (the flow reaches a minimum value of 60% in 10
seconds, versus 25% in the analysis of record). Please provide comparisons between plant
measurements of the pump coastdown curve and the curve assumed in the analysis so that the
NRC staff may verify that appropriately conservative boundary conditions are used. (SRP
15.3.1-15.3.2.III.5: “Time-related variations of the following parameters should be reviewed for
consistency: …core and recirculation loop coolant flow rates”)
Duke Energy RAI 32 Response
The Loss of Forced Reactor Coolant Flow event is defined to result from the instantaneous trip
of one or more reactor coolant pumps. Coolant flow in the core decreases rapidly during the
coastdown of the affected reactor coolant pump(s).
Section 4.3.5 of DPC-NE-3008 compares a RETRAN-3D analysis of the Complete Loss of
Forced Reactor Coolant Flow event to the analysis of record (AOR) for HNP. In the benchmark
analysis, the pump coastdown behavior was closely matched to the AOR to compare the
system thermal-hydraulic response to the loss of flow. In the AOR, the core flow reaches a
value of about 25% in 10 seconds.
Attachment 2 RA-17-0043 Page 38 of 70
Figure RAI-32-1 compares selected flow coastdown results for HNP. In plant measurements of
the flow coastdown, the flow reaches a value of [ ]a,c in 10 seconds. The demonstration
analysis in Section 6.3 of DPC-NE-3009 uses a flow coastdown curve based on the plant
measurements. In the RETRAN-3D model, the pump moment of inertia is adjusted to model a
conservative flow coastdown: the flow reaches a value of 56.6% in 10 seconds. The RETRAN-
3D modeling of the flow coastdown shown in Section 6.3 of DPC-NE-3009 is conservative
relative to the plant measurements, but less restrictive than the AOR treatment. Applications of
DPC-NE-3009 will use a more conservative flow coastdown than the RETRAN-3D modeling
shown in Section 6.3 of DPC-NE-3009.
Attachment 2 RA-17-0043 Page 39 of 70
Figure RAl-32-1 - Loop Flow Coastdown Comparison
a,c
Attachment 2 RA-17-0043 Page 40 of 70
NRC RAI 33
In the locked rotor analysis methodology presented in Section 5.3.2, “Reactor Coolant Pump
Shaft Seizure (Locked Rotor) or Shaft Break,” the use of the VIPRE-01 fuel pin heat conduction
model was proposed for DNBR evaluations. However, no details were provided on how this
model is initialized or what modeling options are used. Is the model used in the same manner
as it is for the rod ejection analysis? More detail is needed for the NRC staff to appropriately
review this application of the model. (SRP 15.0.2)
Duke Energy RAI 33 Response
Section 5.3.2 of DPC-NE-3009-P describes the analysis methodology for the Reactor Coolant
Pump Shaft Seizure (Locked Rotor) or Shaft Break events. The evaluation of the transient
results includes a review of time-related variations in fuel centerline temperature. To facilitate
this review, the VIPRE-01 calculation uses the fuel pin heat conduction model.
[
]a,c
Attachment 2 RA-17-0043 Page 41 of 70
[
]a,c
Attachment 2 RA-17-0043 Page 42 of 70
NRC RAI 34
In the HNP locked rotor demonstration analysis presented in Section 6.4, “Reactor Coolant
Pump Locked Rotor (HNP),” it is stated that “the selection of the affected loop has a negligible
effect on the transient results.” However, the NRC staff did not see a basis provided for this
claim. Duke Energy should provide additional justification, preferably in the form of sensitivity
studies, for the statement that the choice of the affected loop has a negligible effect. (SRP
15.3.1-15.3.2.III.2: “The SAR or DCD must present a quantitative analysis of the most limiting
loss of flow transient.”)
Duke Energy RAI 34 Response
In the demonstration analysis, the affected reactor coolant pump is located in Reactor Coolant
System (RCS) Loop 2, which is connected to the pressurizer surge line (refer to Section 6.4 of
DPC-NE-3009). A sensitivity study was performed with the affected reactor coolant pump in a
loop not connected to the pressurizer surge line: RCS Loop 3. The baseline case trips the
unaffected reactor coolant pumps at the time of turbine trip and more closely matches the
assumptions in the analysis of record, which models a positive moderator temperature
coefficient. The sensitivity case changes the affected RCS loop to RCS Loop 3. The results of
the study are discussed below.
The transient results for core power, core flow, core inlet temperature and core exit pressure are
presented in Figures RAI-34-1 to RAI-34-4. These parameters are used in a transient VIPRE-01
analysis to calculate the minimum DNBR. There are negligible differences in the transient
results for normalized core power (Figure RAI-34-1), and both cases exhibit a power increase
due to the positive moderator temperature coefficient. The transient results for core inlet mass
flux (Figure RAI-34-2) and core inlet temperature (Figure RAI-34-3) show negligible differences
through the time of minimum DNBR. The core exit pressure shows about a 10 psi difference
around the time of minimum DNBR (Figure RAI-34-4), with greater differences in pressure
during the depressurization phase of the transient. The difference in minimum DNBR is
calculated to be 0.001, with minimum DNBR occurring at 3.9 seconds in both cases.
Attachment 2 RA-17-0043 Page 43 of 70
120
100 -ctJ E 80 c ..... 0
60 ~ 0 -... Q)
:= 0
40 0..
20
Figure RAl-34-1 - Normalized Core Power
• • : __......... ......- I\ ' \ • ' • • \ • • ' • ' ' • • • '---• ' ·- -Baseline
• - - Sensitivity '
0 - - - - 1 - _1 ____ - - - - - - - - - - - - - - - - - - - - - -
3.0
-2.5 ¢:: .!. s:. s 2.0 Jl
:i: -1 .5 )( ::::s u.. "' 1.0 "' ctJ
:i: 0.5
0.0
. . 0 1 2 3 4 5
Time (s) 6 7
Figure RAl-34-2 - Core Inlet Mass Flux
\ ' ~- --...... i--.-_ •
..... -Baseline •
- - - - Sensitivity ••• 1 •••• 1 •••• . . . --
0 1 2 3 4
. . . . - . - . 5
Time (s)
. - . -6 7
8 9
- - - ... - - - -8
- - - -
10
.. - - -9 10
Attachment 2 RA-17-0043 Page 44 of 70
Figure RAl-34-3 - Core Inlet Temperature
560
559
558 u:- 557 0 -~ 556 :::J ~ 555 ..... 8. 554 E ~ 553
552
551
550
2,600
2,500
-: 2,400 a. -Q) ; 2,300 in in Q)
! !
~ I~ --
! • ~
·- -Baseline
·- - - Sensitivity ____ 1 ____ 1 ____
- -0 1 2
- - - -3 4
~ ........... ~
- - - . .. - . -5
Time (s) 6
. - . -
Figure RAl-34-4 - Core Exit Pressure
-~
_/ Ii"'""
~ _/
,r ~I\
,, ~
' ~ ....
7
~ ~ - --~
.. - - .. - - - .. - . 8 9 10
n.. 2,200 ~
" 0 ... --"'--2,100 ·- -Baseline
- - Sensitivity 2,000 - - - .L - - - .L - - - - -. .
0 1 2 3 4
- - - - - - -
5 Time (s)
6
- - -
7
- - - - - - - - -
8 9 10
Attachment 2 RA-17-0043 Page 45 of 70
NRC RAI 35
Duke Energy proposed to analyze minimum DNBR using the SCD methodology for the
uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal from subcritical or low
power transient, as described in Section 5.4.1, “Uncontrolled RCCA Bank Withdrawal from a
Subcritical or Low Power Startup Condition.” Considering the power starts at zero or a very low
level for this transient, please justify that the analysis will fall into the range of applicability of the
SCD methodology. What action would be taken if a particular case did not fall within the SCD
methodology range? (SRP 4.4, conformance with the DPC-NE-2005 methodology)
Duke Energy RAI 35 Response
The HNP and RNP SCD ranges of applicability are selected to include the limiting statepoints
from the current analyses of record for the event denoted in DPC-NE-3009 as Uncontrolled
RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition. An event is
applicable to the SCD methodology if the thermal-hydraulic system conditions at the time of
minimum DNBR, defined as the “limiting statepoint,” fall within the SCD range of applicability as
shown in Table H-7 (RNP) and Table I-7 (HNP) of DPC-NE-2005 (Reference RAI-35-1). The
Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition event
is analyzed at hot zero power. Initial core inlet temperature, core exit pressure and core flow at
hot zero power fall within the SCD applicable range and are not expected to exceed the range
at the limiting statepoint. Initial core power is nearly zero, which is outside the SCD range of
applicability. During the power excursion, however, core power typically increases into the SCD
range of applicability. As such, core power is expected to be within the range of applicability of
the SCD methodology at the time of minimum DNBR.
If the limiting statepoint falls outside the SCD range, then either the SCD range of applicability
will be adjusted as described in the response to RAI 15a, or the non-SCD methodology will be
used for the analysis as described in the response to RAI 15b.
Attachment 2 RA-17-0043 Page 46 of 70
Duke Energy RAI 35 Response Reference
RAI-35-1. “Thermal-Hydraulic Statistical Core Design Methodology,” DPC-NE-2005-PA,
Revision 5, March 2016.
Attachment 2 RA-17-0043 Page 47 of 70
NRC RAI 37
The primary difference between the withdrawal of a single full-length RCCA event presented in
Section 5.4.4, “Withdrawal of a Single Full-Length RCCA,” and the uncontrolled RCCA bank
withdrawal at power event presented in Section 5.4.2, “Uncontrolled RCCA Bank Withdrawal at
Power,” is that the former provides a much more localized effect than the latter, resulting in
severe radial power peaking. In the RNP demonstration analysis for this transient, it is unclear
whether these effects are considered. For example, the analysis of record in the RNP UFSAR
predicts DNB in the immediate vicinity of the withdrawn rod due to highly localized peaking,
while the demonstration analysis predicts a minimum DNBR of 1.571. Therefore, please discuss
in more detail how it is ensured that the core power distribution is appropriately calculated using
the methodology discussed in Section 5.4.4, particularly with regard to local power distributions
in the immediate vicinity of the withdrawn control rod. Are pin-level effects considered? (SRP
15.0.2, SRP 15.4.3.III.2 (Reference 19))
NRC RAI 37 Reference
19. U. S. Nuclear Regulatory Commission, NUREG-0800, “Standard Review Plan,” Section
15.4.3, “Control Rod Misoperation (System Malfunction or Operator Error),” Revision 3,
March 2007 (ADAMS Accession No. ML063600415).
Attachment 2 RA-17-0043 Page 48 of 70
Duke Energy RAI 37 Response
The DNB evaluation for the Withdrawal of a Single Full-Length RCCA event consists of two
main parts. The first part determines the system thermal-hydraulic response using RETRAN-3D
and uses [ ]a,c to develop
maximum allowable radial peaking (MARP) limits. The second part generates core power
distributions using SIMULATE-3 for comparison against MARP and centerline fuel melt (CFM)
limits, to determine the fraction of fuel pins that exceed each limit. The SIMULATE-3 power
distributions include the local spatial effects produced from the withdrawal of the control rod.
The cycle-specific core power distributions used in both DNB and CFM evaluations model the
localized peaking increase produced from the withdrawal of the control rod. The analysis
assumes the [
]a,c The failed fuel fractions calculated for each rodded core
configuration are then compared against the values assumed in the dose analysis to ensure the
offsite dose consequences remain within regulatory limits.
The simplified DNB analysis described in Section 6.6 assumed a radial peaking factor of 1.80
and a [ ]a,c. The assumed radial and axial peaking
factors and axial shape do not reflect the local spatial effects from SIMULATE-3. Consequently,
the minimum DNBR calculated in the demonstration analysis is higher than the minimum DNBR
expected in the licensing analysis, and is not intended to imply that the DNBR limit will not be
approached or exceeded for some rod locations in the cycle-specific evaluations.
Attachment 2 RA-17-0043 Page 49 of 70
NRC RAI 38
What reactivity insertion rate was assumed in the withdrawal of a single full-length RCCA
demonstration analysis presented in Section 6.6, “Withdrawal of a Single Full-Length RCCA
(RNP),” of the methodology report? (SRP 15.4.3.III.1)
Duke Energy RAI 38 Response
The reactivity insertion rate assumed in the demonstration analysis presented in Section 6.6
was 0.95 pcm/s. This rate was selected as a reactivity insertion rate sensitivity study showed
that it produced the lowest transient minimum DNBR. The preliminary study examined a
spectrum of reactivity insertion rates from 0.1 pcm/s to 2.35 pcm/s.
Attachment 2 RA-17-0043 Page 50 of 70
NRC RAI 39
Duke Energy stated that the peaking resulting from statically misaligned rods (discussed in
Section 5.4.5, “Static Misalignment of a Single Full-Length RCCA”), is analyzed to confirm that
DNB or CFM will not occur. No transient systems analysis is performed. The licensee stated
that the peaking analysis is a steady-state, three-dimensional power peaking analysis, which
leads the staff to believe that the analysis is performed using SIMULATE-3; however, this was
not specified by the licensee. Please specify the name of the code used for the peaking
analysis? (SRP 15.0.2)
Duke Energy RAI 39 Response
SIMULATE-3 is the current NRC-approved, three-dimensional computer code that will be used
to perform the power peaking analysis for the static misalignment of a single full-length RCCA.
However, the analysis methodology is independent of the model being used to perform the
power peaking analysis. An alternate NRC-approved, three-dimensional computer code could
be used in future reload analyses provided the computer code has received NRC review and
approval. Justification of the computer code’s applicability to the Harris and Robinson cores
would have to be performed through either a license amendment request (LAR) or the Generic
Letter 83-11 process. Additionally, since the computer code would be used to establish core
operating limits, an LAR would be required to add this computer code to the list of NRC-
approved methodologies used to develop core operating limits.
Attachment 2 RA-17-0043 Page 51 of 70
NRC RAI 42
The fuel assembly misloading analysis methodology described in Section 5.4.7, “Inadvertent
Loading of a Fuel Assembly in an Improper Position,” has insufficient detail for the NRC staff to
review it. Though it is apparent from the discussion that part of the evaluation hinges on whether
or not a misloading would be detected during refueling and power ascension operations and
testing, the acceptance criteria used to detect a misloading was not clearly defined.
Furthermore, the method for evaluating DNBR and CFM for misloadings that were not
detectable was not discussed in the report. Please provide the specific criteria used to detect a
misloaded assembly and the method used when misloadings are not detected. (SRP 15.0.2)
Duke Energy RAI 42 Response
A fuel misload may result in a reactivity deviation between the as-loaded core and the reference
core design evaluated in the safety analysis. The primary concern is if the reactivity deviation
results in an increase in peaking factors that can challenge or exceed dose analysis DNB or
CFM failure assumptions. The principal protections against fuel misloads are administrative
procedures employed during the loading of fuel assemblies and components into the reactor
core. During fabrication, each fuel assembly is marked with a unique identification number to
enable easy differentiation among fuel assemblies. Following core loading and prior to
reassembling the reactor vessel, the fuel assembly identification number in each core location is
checked against a core loading diagram to ensure each assembly is in its appropriate location
and orientation.
Attachment 2 RA-17-0043 Page 52 of 70
Should these administrative controls fail, protection is provided through deviations observed
during post-refueling zero-power physics testing and during performance of low- and
intermediate-power flux maps. The zero-power physics testing criteria, such as the comparisons
of the all-rods-out critical boron concentration and bank-specific control rod worths, can detect
gross misloads. However, the principal method to detect fuel misloads is the development and
use of misload screening criteria for the low-power flux map, typically performed near 30%
power. These screening criteria are defined so that any undesirable misloads will be detected.
For this use, an undesirable misload is defined as one which would cause the core to exceed
DNB or CFM failures assumed in the plant-specific dose analysis. Because plant loading
patterns and core power distributions change from cycle to cycle, the screening criteria are by
necessity cycle-specific. In addition, the screening criteria also depend on the number and
location of incore thimbles which are available during the low-power flux map. In general, a flux
map with fewer measured detector locations requires more restrictive screening criteria to
detect undesirable misloads. A flux map which exceeds the screening criteria does not
necessarily indicate a misload, but requires further examination to ensure power distribution
limits will not be exceeded.
The cycle-specific misload analysis calculates three-dimensional power distributions for a
spectrum of potential misload events. [
]a,c The acceptance criterion for the misload analysis is that any misload which
does not exceed the screening criteria (i.e., is undetected) will not invalidate the dose analysis
assumptions on DNB or CFM.
Attachment 2 RA-17-0043 Page 53 of 70
Power distributions for the misload events [
]a,c Since the misloaded fuel assembly event is a Condition III accident,
some limited amount of fuel failures is allowed. If the number of DNB or CFM failures for a given
misload exceeds the dose analysis assumptions, then the screening criteria for the applicable
number of detector location failures must be revised so that case is detected.
An example of cycle-specific misload screening criteria is shown in Table RAI-42-1.
Table RAI-42-1 – Example of Cycle-Specific Misload Screening Criteria for the Initial Low-Power Flux Map
a,c
[
]a,c
Attachment 2 RA-17-0043 Page 54 of 70
NRC RAI 45
In Section 5.4.8.1, “Nuclear Analysis,” Duke Energy states that during the trip following a rod
ejection accident the control rods fall into the core “at a speed that satisfies the maximum rod
drop time in the technical specifications.” Since TSs are derived from the accident analysis,
please specify if the time used in the analysis is consistent with the time specified in the TSs or
is the rod drop time shorter? (SRP 15.4.8.III.1.D (Reference 20))
NRC RAI 45 Reference
20. U. S. Nuclear Regulatory Commission, NUREG-0800, “Standard Review Plan,” Section
15.4.8, “Spectrum of Rod Ejection Accidents (PWR),” Revision 3, March 2007 (ADAMS
Accession No. ML070550014).
Duke Energy RAI 45 Response
The rod drop time used for the Spectrum of RCCA Ejection Accidents event is consistent with
the rod drop times specified in Technical Specifications 3.1.3.4 (HNP) and 3.1.4.3 (RNP).
Technical Specification rod drop times are verified for all control rods through the performance
of rod drop timing tests, prior to unit startup following each refueling outage, to ensure safety
analysis assumptions pertaining to this parameter are satisfied.
Attachment 2 RA-17-0043 Page 55 of 70
NRC RAI 46
In Section 5.4.8.1, Duke Energy stated that the SIMULATE-3K heat conduction model is used to
calculate the temperature distribution within the pin as well as the transport of heat from the fuel,
through the gap and cladding, and into the coolant. The model is also used to [[
]]. It is unclear to the NRC staff whether and how this fuel temperature calculation
feeds back to the neutronics solution. Are individual pin radial temperature distributions
determined and used to calculate Doppler reactivity effects? If average fuel temperatures are
used to evaluate Doppler reactivity effects, please justify why this is appropriate. (SRP 15.0.2)
Duke Energy RAI 46 Response
Individual pin radial temperature distributions are not used to calculate Doppler feedback. All
fuel rods [
]a,c is used for Doppler feedback. The radial
geometry is typically modeled using [ ]a,c radial nodes per fuel assembly, and 24 equal-
length fuel nodes in the axial direction for current designs (Section 5.4.8.1 of DPC-NE-3009).
Calculating the Doppler feedback [
]a,c
both fuel temperatures and enthalpies calculated with this model will be conservative.
Attachment 2 RA-17-0043 Page 56 of 70
Reference 8 of DPC-NE-3009 describes the benchmark of the SIMULATE-3K model against
reference solutions developed for a power ramp, uncontrolled bank withdrawal and control rod
ejection accident. The good agreement between SIMULATE-3K-predicted reactivities relative to
those for the reference solution justifies the acceptability for using the average rod fuel
temperature for calculating Doppler feedback. The good agreement also demonstrates the
fidelity of the neutronics and the coupled transient neutronics and thermal-hydraulic models for
calculating not only Doppler reactivity, but also total core reactivity.
Attachment 2 RA-17-0043 Page 57 of 70
NRC RAI 47
In Section 5.4.8.2, “Fuel Temperature and Enthalpy Calculations,” Duke Energy proposed the
use of [[
]]. Is the [[ ]] incorporated into VIPRE-01? If so, how? If not,
please provide additional justification of how [[
]]. (SRP 4.2, SRP 15.4.8.III.1.E, SRP 15.4.8.III.2.B)
Duke Energy RAI 47 Response
The [ ]a,c has not been incorporated into VIPRE-01. [
]a,c
Attachment 2 RA-17-0043 Page 58 of 70
NRC RAI 49
In Section 5.4.8.4, “RETRAN-3D System Thermal-Hydraulic Calculations,” Duke Energy states
that the failure of the control rod drive mechanism housing that causes the rod ejection is also
assumed to create a hole in the reactor vessel head the size of the control rod drive shaft.
However, the boundary conditions of the hole or how it is modeled are not specified. Please
provide additional information describing how the hole is modeled and the boundary conditions
used. (SRP 15.4.8.III.1.E)
Duke Energy RAI 49 Response
As discussed in Section 5.4.8.4 of DPC-NE-3009, the mechanical failure that results in the
ejected rod is assumed to cause a hole to open in the reactor vessel head. The minimum size of
this hole is assumed to have a diameter equal to the plant-specific diameter of the control rod
drive shaft, because the hole must be large enough to allow the control rod drive shaft to eject
from the reactor vessel. The RETRAN-3D analysis of the Spectrum of RCCA Ejection Accidents
event models the minimum hole size. [
]a,c
Attachment 2 RA-17-0043 Page 59 of 70
NRC RAI 53
In the methodology presented for analysis of the inadvertent operation of the ECCS in Section
5.5.1, “Inadvertent Operation of the Emergency Core Cooling System,” of the report, it is unclear
to the NRC staff whether boron injection is modeled during the transient. The HNP analysis of
record for this event explicitly considers the effects of boric acid on reactivity. If boron injection is
modeled in the proposed methodology, please discuss the assumptions used related to boron
concentrations, boron worth, and boron transport. If not, please discuss how the proposed
method appropriately accounts for the effects of inadvertent operation of the ECCS on the core
power level. (Consistency with HNP current licensing basis (CLB))
Duke Energy RAI 53 Response
Boron injection is modeled. The model used is the General Transport Model described in
Section 5.1.4.1 of DPC-NE-3009. Maximum boron concentration and boron worth are assumed
to reduce core power, leading to reduced RCS pressure and coolant temperatures. Minimum
boron transport delays (flushing volume) are assumed to minimize the core power response.
Attachment 2 RA-17-0043 Page 60 of 70
NRC RAI 54
Section 5.6.1, “Inadvertent Opening of a Pressurizer Relief or Safety Valve,” of the methodology
presents the proposed methodology for analyzing the inadvertent opening of a pressurizer relief
or safety valve. It is unclear to the staff whether both power-operated relief valves and safety
valves are considered as potential failures, or if the limiting valve is selected in advance. It is
also unclear how the initiating valve failure is modeled. Please clarify. (SRP 15.6.1 (Reference
21))
NRC RAI 54 Reference
21. U. S. Nuclear Regulatory Commission, NUREG-0800, “Standard Review Plan,” Section
15.6.1, “Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR
Pressure Relief Valve,” Revision 2, March 2007 (ADAMS Accession No. ML070820094).
Duke Energy RAI 54 Response
The analysis is applicable to either valve, but since the capacity of a safety valve bounds that of
a PORV, a stuck-open safety valve is chosen. The valve is modeled assuming the maximum
flow capacity of the valve to maximize the depressurization and cooldown. The critical flow
model used in the analysis is the combination model comprised of the Extended Henry
(subcooled) and Moody (saturated) models. The effective area of the safety valve is adjusted to
achieve the maximum flow capacity of the valve at the rated conditions. The failure is modeled
as a rapid opening of the safety valve to the fully open position.
Attachment 2 RA-17-0043 Page 61 of 70
NRC RAI 55
The SCD methodology is used to evaluate the core cooling capability for the inadvertent
opening of a pressurizer relief or safety valve transient in Section 5.6.1. Given that this event
has the potential for core uncovery, please justify why it is appropriate to evaluate the event
from nominal, rather than conservative initial conditions. Is a separate core uncovery analysis
performed? (SRP 15.0.2, SRP 15.6.1.I.1.B)
Duke Energy RAI 55 Response
This event presents two possible challenges to core cooling. The first challenge is near the time
of reactor scram caused by the decrease in RCS pressure. Core uncovery is not a concern
during this short time period. This phase of the analysis is performed using the SCD
methodology assuming nominal initial conditions. The long-term phase of this event is
addressed as part of the LOCA analysis for Robinson assuming bounding initial conditions.
(Note that a failure of a PORV or safety valve is classified as a Condition IV event for
Robinson.) For Harris, the long-term response was addressed as part of the Steam Generator
Replacement / Power Uprate project, and AREVA (Siemens) determined using the SBLOCA
analysis methods that core cooling capability is adequate. The analysis is performed with
conservative initial conditions (e.g., power level, temperature, flow rate), and no core uncovery
or DNB occurs. The long-term portion of these analyses, should they need to be reanalyzed,
would be performed using the SBLOCA analysis methods and would assume conservative
initial conditions.
Attachment 2 RA-17-0043 Page 62 of 70
NRC RAI 56
For the steam generator tube rupture, the licensing basis analysis presented in the HNP FSAR
focuses almost exclusively on the margin to steam generator overfill, which is not presented as
an acceptance criterion in Section 5.6.2, “Steam Generator Tube Rupture,” of the methodology
report. Please discuss whether steam generator overfill will be analyzed with the proposed
method, with a justification for why the analysis is or is not needed. This discussion should
consider single failure assumptions for the margin to overfill analysis, consistent with the HNP
FSAR. (Consistency with HNP CLB)
Duke Energy RAI 56 Response
The Margin to Overfill analysis will be performed with the proposed method. HNP is a Standard
Review Plan plant and must demonstrate that the steam generator (SG) will not overfill following
a steam generator tube rupture (SGTR). The HNP FSAR also states that once Margin to Overfill
is demonstrated, an Offsite Dose Assessment can be performed which has different sensitivities
to plant parameters. Therefore, both the Margin to Overfill and Offsite Dose analyses will be
performed with the proposed method. The Offsite Dose analysis described below is only the
thermal-hydraulic portion and does not address any radiological portion of the analysis.
Margin to Overfill Analysis
The SGTR overfill methodology is largely based on the PWR Owners’ Group work described in
WCAP-10698-P-A (Reference RAI-56-1). Following is a summary of the initial conditions,
boundary conditions, single failures and operator actions that are significant to the analysis.
Attachment 2 RA-17-0043 Page 63 of 70
Initial Conditions
All initial conditions significant to the analysis are listed below. Any initial conditions not listed
below are assumed to be nominal values.
● RCS Tavg: Sensitivity studies will be performed to determine what is limiting.
● SG Tube Plugging: Sensitivity studies will be performed to determine what is limiting.
● Power: Full power plus uncertainty.
● Pressurizer Pressure: Sensitivity studies will be performed to determine what is limiting.
● SG Level: High (Nominal + uncertainty + 10% to accommodate turbine runback).
Boundary Conditions
All boundary conditions significant to the analysis are listed below.
● Reactor Trip Signal Expected On:
– Low Pressurizer Pressure: Nominal + uncertainty (high is conservative)
– Over-Temperature Differential Temperature (OTΔT)
● Loss of offsite power upon reactor trip.
● Low Pressurizer Pressure Safety Injection: Nominal + uncertainty (high is conservative).
● Break location is at the top of the tube sheet on the cold side of the SG, to maximize the
primary-to-secondary flow rate.
● Maximum AFW flow with initiation of both motor-driven AFW pumps and the turbine-driven
AFW pump. Minimum temperatures will be applied to AFW flow.
● Safety injection flow from two high head safety injection pumps.
● Decay Heat: Sensitivity studies will be performed to determine if high or low decay heat is
limiting.
Attachment 2 RA-17-0043 Page 64 of 70
Single Failures
The following single failures will be evaluated to determine which is the most limiting:
● Failure of an intact SG PORV to open
● Failure of the TDAFW Pump Speed Controller
● Failure of an AFW flow control valve to close on demand
Operator Actions
The following operator actions are modeled. These are the same operator actions currently
credited in FSAR Section 15.6.3.
● Identify the ruptured SG and control AFW flow 8.8 minutes from event initiation or when
narrow range level reaches 30% on the ruptured SG, whichever is longer.
– In the case of the assumed failure of the AFW flow control valve to close, an additional 2
minute delay is applied to allow time for operators to close the AFW isolation valve.
● Isolate the ruptured SG 12 minutes from event initiation.
● Initiate RCS cooldown with the intact SG PORVs 5 minutes following isolation of the
ruptured SG. Cool down the RCS to 20°F subcooling.
● Initiate depressurization with a pressurizer PORV 4 minutes after termination of RCS
cooldown.
● Terminate SI 3 minutes after the end of the depressurization.
Offsite Dose Thermal-Hydraulic Input Analysis
The SGTR offsite dose methodology is largely based on the PWR Owners’ Group work
described in WCAP-10698-P-A. Following is a summary of the initial conditions, boundary
conditions, single failures and operator actions that are significant to the analysis.
Attachment 2 RA-17-0043 Page 65 of 70
Initial Conditions
All initial conditions significant to the analysis are listed below. Any initial conditions not listed
below are assumed to be nominal values.
● RCS Tavg: Sensitivity studies will be performed to determine what is limiting.
● SG Tube Plugging: Sensitivity studies will be performed to determine what is limiting.
● Power: Full power plus uncertainty.
● Pressurizer Pressure: Sensitivity studies will be performed to determine what is limiting.
Boundary Conditions
All boundary conditions significant to the analysis are listed below.
● Reactor Trip Signal Expected On:
– Low Pressurizer Pressure: Nominal - uncertainty (low is conservative)
– Over-Temperature Differential Temperature (OTΔT)
● Loss of offsite power upon reactor trip.
● Break location is at the top of the tube sheet on the cold side of the SG, to maximize the
primary-to-secondary flow rate.
● Decay Heat: High decay heat is conservative in order to maximize the post-trip heatup of the
primary system.
Single Failure
● The SG PORV on the ruptured SG fails to close at the time that operators isolate the
ruptured SG.
Attachment 2 RA-17-0043 Page 66 of 70
Operator Actions
The following operator actions are modeled. These are the same operator actions currently
credited in FSAR Section 15.6.3.
● Identify the ruptured SG and control AFW flow 10 minutes from event initiation or when
narrow range level reaches 30% on the ruptured SG, whichever is longer.
● Isolate the ruptured SG 12 minutes from event initiation.
– Ruptured SG PORV fails to close.
● Operators have identified the failed-open SG PORV on the ruptured SG and locally close
the associated block valve 20 minutes following isolation of the ruptured SG.
● Initiate RCS cooldown with the intact SG PORVs 5 minutes following closure of the block
valve. Cool down the RCS to 20°F subcooling.
● Initiate depressurization with a pressurizer PORV 4 minutes after termination of RCS
cooldown.
● Terminate SI 3 minutes after the end of the depressurization.
RETRAN-3D Safety Evaluation Report Limitations and Conditions
The limitations and conditions from the NRC’s generic Safety Evaluation Report on the
RETRAN-3D computer code (Reference RAI-56-2) were reviewed for the application of the
HNP/RNP base models to the Margin to Overfill and Offsite Dose Thermal-Hydraulic Input
analyses. The limitations and conditions are addressed by previous statements of resolution in
Section 3.2 of DPC-NE-3009.
Attachment 2 RA-17-0043 Page 67 of 70
Duke Energy RAI 56 Response References
RAI-56-1. “SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,”
WCAP-10698-P-A, August 1987.
RAI-56-2. Richards, S. A., to Vine, G. L., “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),” January 25, 2001.
(ADAMS Accession No. ML010470342)
Attachment 2 RA-17-0043 Page 68 of 70
NRC RAI 57
The current licensing basis for HNP relies on several operator actions to recover from a steam
generator tube rupture. Please discuss whether operator actions are modeled to terminate the
steam generator tube rupture in the DPC-NE-3009 methodology. If operator actions are
required, please describe and justify how each operator action is modeled. (There is no SRP
section specifically for the review of steam generator tube ruptures, but the requirement to
identify operator actions credited in transient and accident analyses is in SRP 15.0.I.6)
Duke Energy RAI 57 Response
The description and justification of how each operator action is modeled in the SGTR Margin to
Overfill and Offsite Dose Thermal-Hydraulic Input analyses is discussed below. The modeled
operator actions mimic real operator actions performed in the SGTR Emergency Operating
Procedure.
SGTR Margin To Overfill Analysis Operator Actions
● Identify the ruptured SG and control AFW flow 8.8 minutes from event initiation or when
narrow range level reaches 30% on the ruptured SG, whichever is longer.
– AFW flow begins on reactor trip at maximum flow until 8.8 minutes. At this point, a
control system will throttle AFW flow to a fixed level setpoint in the ruptured SG, which
mimics real operator actions per Emergency Operating Procedures. The narrow range
level of 30% is a reasonable level setpoint that matches WCAP-10698-P-A (Reference
RAI-57-1).
● Isolate the ruptured SG 12 minutes from event initiation.
– All ruptured SG valves, such as the MSIV, are closed, isolating the ruptured SG. This
mimics real operator actions per Emergency Operating Procedures.
Attachment 2 RA-17-0043 Page 69 of 70
● Initiate RCS cooldown with the intact SG PORVs 5 minutes following isolation of the
ruptured SG. Cool down the RCS to 20°F subcooling.
– SG PORVs on the intact SGs are opened to cool down the RCS until the correct core
exit temperatures from Emergency Operating Procedure E-3 are obtained. The target
core exit temperatures are based on 20°F RCS subcooling.
● Initiate depressurization with a pressurizer PORV 4 minutes after termination of RCS
cooldown.
– Normal methods of depressurizing the RCS such as pressurizer spray are unavailable
due to the loss of offsite power. Therefore, a pressurizer PORV is used to depressurize
the RCS until one of the following criteria is met: 1) RCS pressure is less than the
ruptured SG pressure; 2) pressurizer level increases above the level specified in the
Emergency Operating Procedures; or 3) RCS subcooling decreases below 20°F.
● Terminate SI 3 minutes after the end of the depressurization.
– Safety injection flows are terminated, which mimics real operator actions per Emergency
Operating Procedures.
SGTR Offsite Dose Thermal-Hydraulic Input Analysis Operator Actions
● Identify the ruptured SG and control AFW flow 10 minutes from event initiation or when
narrow range level reaches 30% on the ruptured SG, whichever is longer.
– AFW flow begins on reactor trip at maximum flow until 10 minutes. At this point, a control
system will throttle AFW flow to a fixed level setpoint in the ruptured SG, which mimics
real operator actions per Emergency Operating Procedures. The narrow range level of
30% is a reasonable level setpoint similar to the modeling in WCAP-10698-P-A.
● Isolate the ruptured SG 12 minutes after event initiation.
– All valves, such as the MSIV, to the ruptured SG are closed, hydraulically isolating the
ruptured SG from the rest of the system. This mimics real operator actions per
Emergency Operating Procedures.
Attachment 2 RA-17-0043 Page 70 of 70
● Operators have identified the failed-open SG PORV on the ruptured SG and locally close
the associated block valve 20 minutes following isolation of the ruptured SG.
– This mimics real operator actions per Emergency Operating Procedures.
● Initiate RCS cooldown with the intact SG PORVs 5 minutes following closure of the block
valve. Cool down the RCS to 20°F subcooling.
– SG PORVs on intact SGs are opened to cool down the RCS until the correct core exit
temperatures from Emergency Operating Procedure E-3 are obtained. The target core
exit temperatures are based on 20°F RCS subcooling.
● Initiate depressurization with a pressurizer PORV 4 minutes after termination of RCS
cooldown.
– Normal methods of depressurizing the RCS such as pressurizer spray are unavailable
due to the loss of offsite power. Therefore, a pressurizer PORV is used to depressurize
the RCS until one of the following criteria is met: 1) RCS pressure is less than the
ruptured SG pressure; 2) pressurizer level increases above the level specified in the
Emergency Operating Procedures; or 3) RCS subcooling decreases below 20°F.
● Terminate SI 3 minutes after the end of the depressurization.
– Safety injection flows are terminated, which mimics real operator actions per Emergency
Operating Procedures.
Duke Energy RAI 57 Response Reference
RAI-57-1. “SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,”
WCAP-10698-P-A, August 1987.