ML18323A269
| ML18323A269 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 11/19/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-1118-62982 | |
| Download: ML18323A269 (13) | |
Text
RAIO-1118-62982 November 19, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Supplemental Response to NRC Request for Additional Information No. 256 (eRAI No. 9161) on the NuScale Design Certification Application
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 256 (eRAI No. 9161 )," dated October 13, 2017
- 2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 256 (eRAI No.9161)," dated August 30, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).
The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9161:
- 11.01-1 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions on this response, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.
Sincerely,
~
/Zackary W. Rad
~
Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Getachew Tesfaye, NRC, OWFN-8H12 Enclosure 1: NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9161 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com
NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9161 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
Response to Request for Additional Information Docket No.52-048 eRAI No.: 9161 Date of RAI Issue: 10/13/2017 NRC Question No.: 11.01-1 Regulatory Requirements/Guidance 10 CFR 52.47(a)(5) requires applicants to identify the kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radiation exposures. Radioactive materials are released as liquid and gaseous effluents and are generated in plant systems during normal reactor operations resulting in doses to workers and members of the public. 10 CFR Part 20, Appendix I to 10 CFR Part 50, and 40 CFR Part 190 specify the annual dose limits to workers and members of the public, and the As Low As is Reasonably Achievable numerical objectives in the design of radwaste systems for controlling and limiting liquid and gaseous effluent releases.
The Design Specific Review Standard (DSRS) Acceptance Criteria section for NuScale, DSRS Section 11.1, Coolant Source Terms, states when the applicants calculation technique or any source term parameters differ from that given in NUREG-0017 or ANSI/ANS 18.1-1999, they should be described with sufficient detail, and the basis of the alternate method and model parameters should be provided to allow the staff to conduct an independent evaluation. DSRS Sections 11.2, Liquid Waste Management System, and 11.3, Gaseous Waste Management System, describe that the calculated annual total quantity of radioactive materials released from each reactor will not result in exceeding the annual exposure pathway doses from liquid and gaseous effluents in Appendix I to 10 CFR Part 50; annual dose limits in 10 CFR 20.1301; and annual liquid and gaseous effluent concentration limits in Table 2, Columns 1 and 2 of Appendix B to 10 CFR Part 20. Further, DSRS Section 12.2, Radiation Sources, states that applications should contain the methods, models, and assumptions used as the bases for all sources described.
NuScale Nonproprietary
Key Issue: NuScale is pursuing an alternate approach for developing realistic liquid and gaseous effluent release source terms during normal operations from that endorsed by NRC in its current guidance. Any differences from the endorsed approach must be identified and justified.
Design certification application content:
NuScale Technical Report TR-1116-52065-NP Rev. 0, Effluent Release (GALE Replacement)
Methodology and Results, provides a proposed alternative methodology for determining the realistic fuel failure fraction of 0.000028 (0.0028 percent (%) or 28 parts per million (ppm))
applied in fundamental first principle calculations of liquid and gaseous effluent source terms expected from normal operating conditions for the NuScale design. This realistic fuel failure fraction is based on a literature search as there is no operating experience or data available for performance of the shorter-in-length AREVA M5 fuel in a first-of-a-kind small modular reactor design. From the staffs review of the literature referenced in TR-1116-52065-NP Rev. 0; Chapter 11, Radioactive Waste Management of the Design Control Document (DCD); and information obtained independently by the staff, there is an insufficient technical bases and justification to support the realistic failed fuel fraction in TR-1116-52065-NP Rev. 0.
The staff requests that NuScale provide the following information for its review of the realistic failed fuel fraction to evaluate compliance with the applicable NRC requirements:
- 1. NuScale needs to provide a technical basis that clearly identifies and describes the fuel failure methodology (e.g., how fuel failure rate is evaluated using an approach such as the outage method). As part of its evaluation of the technical bases, the staff needs to review and confirm key parameters, values and assumptions used in calculating a conservative realistic failed fuel fraction appropriate for determining reactor coolant system (RCS) activity concentrations. Examples of parameters and values include the number of U.S. PWRs determined to be representative, core/assembly size, number of failed pins/rods, number of pins/rods and assemblies in refueling cycle, and total number of rods in the entire core that have been refueled in that year for the range of years considered. Examples of assumptions used in calculating a conservative realistic failed fuel fraction may include a maximum value or average value plus two sigma that bounds the realistic failed fuel fraction. The staff also needs to be provided for its review and evaluation the Electric Power Research Institute (EPRI) source data, other associated information, and quality documentation supporting the calculation of the realistic failed fuel fraction basis.
NuScale Nonproprietary
- 2. In NuScale Technical Report TR-1116-52065-NP Rev. 0, International Atomic Energy Agency, Review of Fuel Failures in Water Cooled Reactors, IAEA Nuclear Energy Series No. NF-T-2.1, June 2010 (Reference 7.2.9) states that there are different methods of fuel failure rate evaluation used by different utilities, fuel vendors and organizations make fuel failure analysis and identification of general tendencies in fuel performance evaluation difficult. It further describes that EPRI uses the fuel assembly failure rate method which may underestimate the defect rate by a factor of about three to five. For example, Table 3.3 in IAEA NF-T-2.1 shows that for U.S. PWRs from 1994 through 2006 there is a factor of 5 underestimation (non-conservativism) in comparing old (core) method of 25.8 parts per million (ppm) and new (reload) method of 131.6 ppm for fuel failure rates.
Provide sufficient detail to identify and justify the failed fuel methodology (i.e., fuel failure rate evaluation) applied in determining the RCS activity concentrations for the NuScale design.
- 3. NuScale Technical Report TR-1116-52065-NP Rev. 0, Section 5.2, US Pressurized Water Reactor Fuel Failure History discusses fuel failure fractions and suggests a correlation to ANSI/ANS-18.1- 1999, NUREG-0017 Rev. 1, PWR-GALE08, and PWR-GALE09.
The ANSI/ANS-18.1-1999 voluntary standard (Reference 7.2.2) and NUREG-0017 Rev. 1 guidance (Reference 7.2.1) do not specify any failed fuel fraction, fuel failure methodology, or discuss the representativeness and quality of these data. Moreover, interim versions of PWR-GALE08 and PWR- GALE09 in Geelhood, K.J. and J.P. Rishel, Applicability of GALE-86 Codes to Integral Pressurized Water Reactor Designs, PNNL-21386, May 2012 (Reference 7.2.3), and Geelhood, K.J., Benchmarking of GALE-09 Release Predictions Using Site Specific Data from 2005 to 2010, PNNL- 22076, November 2012 (Reference 7.2.19) are not endorsed by the NRC.
Given that ANSI/ANS-18.1-1999 and NUREG-0017 Rev. 1 do not provide sufficient information to support a realistic failed fuel fraction basis, and that PWR-GALE08 and PWR-GALE09 are not endorsed by the NRC, provide sufficient detail and justification of how fuel failure rates can be correlated to ANSI/ANS-18.1-1999, NUREG-0017 Rev. 1, PWR-GALE08 and PWR-GALE09, and are appropriate for determining RCS activity concentrations for the NuScale design.
- 4. NuScale Technical Report TR-1116-52065-NP Rev. 0, Appendix A, Table A-6, Fuel failure data for U.S. PWRs with zirconium-alloy cladding provides failed fuel fractions for Sources NuScale Nonproprietary
and References from the early 1970s through 2010. Figure 5-2, Fuel failure data for zirconium-alloy clad U.S. pressurized water reactors plots fuel failure fractions for zirconium-alloy clad U.S. PWRs from 1975 through 2010 from various sources. Table 5-1, Fuel failure values shows Minimum (Rods/Million), Maximum (Rods/Million), and Average (Rods/Million) values for three Date Ranges (1996-2000, 2001-2005, 2006-2010), in which the realistic fuel failure fraction of 0.000028 (0.0028% or 28 ppm) is shown as the maximum value for date ranges 2001-2005.
The NuScale realistic fuel failure fraction of 0.000028 (0.0028% or 28 ppm) is determined from EPRI source data in 2001 and Executive Summary - The Path to Zero Defects: EPRI Fuel Reliability Guidelines, 2008 (Reference 7.2.13). This two page Executive Summary contains a figure (bar graph) showing EOC (end of cycle) from 1980 through 2007 on the x-axis, but there is no unit label on the y-axis to identity what the values ranging from 0 to 300 represent.
- a. Provide the y-axis unit label (e.g., number of failed pins/rods identified in refueling cycle for that year;
- b. Identify the fuel failure methodology (e.g., core, outage, or reload) applied for each year from 1980 through 2007;
- c. Provide the EPRI source data, associated information, and quality documentation used to extrapolate the maximum value (around 95 with no expressed unit label) in 2001 for U.S. PWRs in calculating the realistic fuel failure fraction; and
- d. Provide sufficient detail and justification of how the tables, figures, and fuel failure methodology applied establishes an appropriate basis for determining RCS activity concentrations for the NuScale design.
- 5. In NuScale Technical Report TR-1116-52065-NP Rev. 0, the staff notes that conference presentations such as the International Atomic Energy Agency, Results of the IAEA Study of Fuel Failures in Water Cooled Reactors in 2006-2010, Presentation at Technical Working Group on Fuel Performance and Technology (TWGFPT) Meeting, April 24, 2012 (Reference 7.2.14); and U.S. Nuclear Regulatory Commission, Nuclear Fuel Performance, Office of NuScale Nonproprietary
Nuclear Regulatory Research and Office of Nuclear Reactor Regulation Presentation, February 24, 2005 (Reference 7.2.18) are intended for information purposes only.
Provide sufficient detail and justification of how this information (conference presentations) can be used as the basis for licensing and for determining RCS activity concentrations for the NuScale design.
- 6. NuScale needs to revise the relevant sections, tables, and figures in DCD Chapters 11 and 12 and NuScale Technical Report TR-1116-52065-NP Rev. 0 related to the realistic failed fuel fraction discussions, and provide markups to include this revised information.
NuScale Response:
During a public NRC clarification call on 10/18/2018, NuScale agreed to revise the description of the primary coolant source term, as contained in FSAR Section 12.2.1.2, Table 12.2-1, Table 12.2-3, and Table 12.2-4. This revision is intended to clarify the two regions modeled for the primary coolant - hot leg and cold leg regions. The hot leg region consists of the lower and upper riser areas, and represents 27% of the primary coolant's total mass, and is assumed to contain the core exit N-16 concentration throughout. The cold leg region consists of the pressurizer liquid, the steam generator, and the reactor coolant downcomer regions, and represents 73% of the primary coolant's total mass, and is assumed to contain an N-16 concentration equal to that of the steam generator's entrance throughout. The N-16 concentrations are provided in FSAR Table 12.2-5.
Impact on DCA:
FSAR Section 12.2.1.2, Table 12.2-1, 12.2-3, and 12.2-4 have been revised as described in the response above and as shown in the markup provided in this response.
NuScale Nonproprietary
NuScale Final Safety Analysis Report Radiation Sources RAI 11.01-1S1, RAI 12.02-24S1 The contribution of gamma radiation from the primary coolant is comprised of two components: the hot leg (core, lower riser, and upper riser) and a cold leg (pressurizer, steam generator, and reactor coolant downcomer region). The hot leg is modeled with the peak N-16 concentration in the lower riser, while the cold leg is modeled with an N-16 concentration equal to the entrance of the steam generator. Because of the low flow velocity of the primary coolant, the short half life of N-16 causes it to decay by about one order of magnitude by the time it reaches the steam generator entrance. By uniformly treating the hot leg with the peak N-16 concentration in the RCS loop, and the cold leg uniformly with the steam generator entrance concentration, the gamma contribution from N-16 is conservatively modeled. The fission isotopes and corrosion isotopes (CRUD) are uniformly modeled on a primary coolant mass basis. The primary coolant gamma spectra are provided in Table 12.2-3 and Table 12.2-4.
Nitrogen-16 is present throughout the primary coolant loop, and the modeling simplification described above is conservative from a bioshield radiation shielding perspective. Table 12.2-5 tabulates the nitrogen-16 concentration at several locations in the primary coolant system.
12.2.1.3 Chemical and Volume Control System RAI 12.02-14 The chemical and volume control system (CVCS) takes a portion of the RCS and processes the water through heat exchangers, demineralizers, and filters. The treated primary coolant water is then returned to the RCS (Section 9.3.4). During this treatment process, components of the CVCS can become radiation sources due to soluble and non-soluble radionuclides in the primary coolant. The CVCS contained sources are determined using the design basis coolant source term from Section 11.1 (Table 11.1-4).
Mixed-Bed and Cation Bed Demineralizers RAI 12.02-14 The CVCS mixed-bed demineralizers are assumed to be in continuous operation during the entire fuel cycle. The decontamination factors assumed are listed in Table 11.1-2.
RAI 12.02-14 The CVCS cation bed demineralizers are assumed to be in operation for one-half of the fuel cycle because they are operated intermittently during the operating cycle for lithium removal. The decontamination factors assumed are listed in Table 12.2-6.
The CVCS demineralizer beds are located in the Reactor Building (RXB) on the 24' elevation inside the CVCS cubicles. The mixed-bed source terms and source strengths are listed in Table 12.2-7 and Table 12.2-8, respectively.
Regenerative and Non-Regenerative Heat Exchangers Tier 2 12.2-2 Draft Revision 3
NuScale Final Safety Analysis Report Radiation Sources RAI 11.01-1, RAI 11.01-1S1, RAI 12.02-24S1, RAI 12.03-28 Table 12.2-1: Core and Coolant Source Information Parameter Value Fission neutron source strength 1.91E+19 particles/sec Fission neutron energy spectrum Watt spectrum for U-235 Fission gamma source strength 1.45E+20 particles/sec Near CoreHot Leg Fraction of Primary coolant 27%
Cold Leg Fraction of Primary coolant 73%
Total Primary Coolant source strength 1.06E+14 particles/sec Tier 2 12.2-11 Draft Revision 3
NuScale Final Safety Analysis Report Radiation Sources RAI 11.01-1, RAI 11.01-1S1, RAI 12.02-24S1 Table 12.2-3: Cold Leg Primary Coolant Gamma Source Term Gamma Source Lower Bound Upper Bound (photons/sec/gram primary Energy Group (MeV) (MeV) coolant) 1 1.00E-02 2.00E-02 3.38E+04 2 2.00E-02 3.00E-02 2.50E+04 3 3.00E-02 4.50E-02 1.03E+05 4 4.50E-02 6.00E-02 1.26E+04 5 6.00E-02 7.00E-02 6.19E+03 6 7.00E-02 7.50E-02 2.69E+03 7 7.50E-02 1.00E-01 8.49E+04 8 1.00E-01 1.50E-01 1.32E+04 9 1.50E-01 2.00E-01 1.09E+04 10 2.00E-01 2.60E-01 1.30E+04 11 2.60E-01 3.00E-01 9.78E+03 12 3.00E-01 4.00E-01 9.74E+04 13 4.00E-01 4.50E-01 3.43E+03 14 4.50E-01 5.10E-01 3.59E+03 15 5.10E-01 5.12E-01 1.37E+03 16 5.12E-01 6.00E-01 4.21E+04 17 6.00E-01 7.00E-01 1.33E+04 18 7.00E-01 8.00E-01 6.31E+03 19 8.00E-01 9.00E-01 5.59E+03 20 9.00E-01 1.00E+00 1.61E+03 21 1.00E+00 1.20E+00 3.06E+03 22 1.20E+00 1.33E+00 1.11E+04 23 1.33E+00 1.44E+00 1.46E+03 24 1.44E+00 1.50E+00 3.57E+02 25 1.50E+00 1.57E+00 4.26E+02 26 1.57E+00 1.66E+00 4.35E+02 27 1.66E+00 1.80E+00 1.50E+03 28 1.80E+00 2.00E+00 1.28E+03 29 2.00E+00 2.15E+00 4.89E+02 30 2.15E+00 2.35E+00 3.02E+02 31 2.35E+00 2.50E+00 1.16E+03 32 2.50E+00 2.75E+00 5.03E+03 33 2.75E+00 3.00E+00 1.34E+03 34 3.00E+00 3.50E+00 3.30E+02 35 3.50E+00 4.00E+00 3.71E+02 36 4.00E+00 4.50E+00 1.34E+02 37 4.50E+00 5.00E+00 1.25E+02 38 5.00E+00 5.50E+00 1.11E+02 39 5.50E+00 6.00E+00 5.49E+01 40 6.00E+00 6.50E+00 3.62E+05 41 6.50E+00 7.00E+00 2.37E+02 42 7.00E+00 7.50E+00 2.65E+04 43 7.50E+00 8.00E+00 7.58E+00 44 8.00E+00 1.00E+01 4.23E+02 Tier 2 12.2-14 Draft Revision 3
NuScale Final Safety Analysis Report Radiation Sources Table 12.2-3: Cold Leg Primary Coolant Gamma Source Term (Continued)
Gamma Source Lower Bound Upper Bound (photons/sec/gram primary Energy Group (MeV) (MeV) coolant) 45 1.00E+01 1.20E+01 1.08E-01 46 1.20E+01 1.40E+01 -
47 1.40E+01 2.00E+01 -
Total 9.08E+05 Note: This source term is used for 73 percent by mass of the primary coolant in the NuScale operating reactor shielding calculation, for the pressurizer region above the pressurizer plate and from the steam generator to the cold inlet of the core.
Tier 2 12.2-15 Draft Revision 3
NuScale Final Safety Analysis Report Radiation Sources RAI 11.01-1, RAI 11.01-1S1, RAI 12.02-24S1 Table 12.2-4: Near CoreHot Leg Primary Coolant Gamma Source Term Gamma Source Lower Bound Upper Bound (photons/sec/gram primary Energy Group (MeV) (MeV) coolant) 1 1.00E-02 2.00E-02 2.95E+05 2 2.00E-02 3.00E-02 1.77E+05 3 3.00E-02 4.50E-02 2.28E+05 4 4.50E-02 6.00E-02 1.12E+05 5 6.00E-02 7.00E-02 5.52E+04 6 7.00E-02 7.50E-02 2.41E+04 7 7.50E-02 1.00E-01 1.75E+05 8 1.00E-01 1.50E-01 1.14E+05 9 1.50E-01 2.00E-01 8.99E+04 10 2.00E-01 2.60E-01 6.06E+04 11 2.60E-01 3.00E-01 3.33E+04 12 3.00E-01 4.00E-01 1.61E+05 13 4.00E-01 4.50E-01 2.59E+04 14 4.50E-01 5.10E-01 2.80E+04 15 5.10E-01 5.12E-01 1.37E+03 16 5.12E-01 6.00E-01 5.44E+04 17 6.00E-01 7.00E-01 3.32E+04 18 7.00E-01 8.00E-01 2.17E+04 19 8.00E-01 9.00E-01 1.78E+04 20 9.00E-01 1.00E+00 1.17E+04 21 1.00E+00 1.20E+00 1.80E+04 22 1.20E+00 1.33E+00 1.95E+04 23 1.33E+00 1.44E+00 6.06E+03 24 1.44E+00 1.50E+00 2.36E+03 25 1.50E+00 1.57E+00 2.35E+03 26 1.57E+00 1.66E+00 3.82E+03 27 1.66E+00 1.80E+00 1.14E+04 28 1.80E+00 2.00E+00 9.30E+03 29 2.00E+00 2.15E+00 2.94E+03 30 2.15E+00 2.35E+00 3.02E+02 31 2.35E+00 2.50E+00 7.19E+03 32 2.50E+00 2.75E+00 4.43E+04 33 2.75E+00 3.00E+00 1.11E+04 34 3.00E+00 3.50E+00 2.97E+03 35 3.50E+00 4.00E+00 3.42E+03 36 4.00E+00 4.50E+00 1.24E+03 37 4.50E+00 5.00E+00 1.15E+03 38 5.00E+00 5.50E+00 1.04E+03 39 5.50E+00 6.00E+00 5.12E+02 40 6.00E+00 6.50E+00 3.38E+06 41 6.50E+00 7.00E+00 2.21E+03 42 7.00E+00 7.50E+00 2.47E+05 43 7.50E+00 8.00E+00 7.07E+01 44 8.00E+00 1.00E+01 3.94E+03 Tier 2 12.2-16 Draft Revision 3
NuScale Final Safety Analysis Report Radiation Sources Table 12.2-4: Near CoreHot Leg Primary Coolant Gamma Source Term (Continued)
Gamma Source Lower Bound Upper Bound (photons/sec/gram primary Energy Group (MeV) (MeV) coolant) 45 1.00E+01 1.20E+01 1.01E+00 46 1.20E+01 1.40E+01 -
47 1.40E+01 2.00E+01 -
Total 5.50E+06 Note: This source term is used for 27 percent by mass of the primary coolant in the NuScale operating reactor shielding calculation, for primary coolant leaving the core to the top of the upper riser.
Tier 2 12.2-17 Draft Revision 3