ML18067A570

LLC - Response to NRC Request for Additional Information No. 333 (Erai No. 9282) on the NuScale Design Certification Application

ML18067A570
Person / Time
Site:	NuScale
Issue date:	03/08/2018
From:	Wike J NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0318-59008
Download: ML18067A570 (5)

Text

RAIO-0318-59008 March 08, 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 Response to NRC Request for Additional Information No.

333 (eRAI No. 9282) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

333 (eRAI No. 9282)," dated January 09, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's response to the following RAI Question from NRC eRAI No. 9282:

03.11-17 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 Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

Sincerely, Jennie Wike Manager, Licensing NuScale Power, LLC Distribution: Omid Tabatabai, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9282 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0318-59008 :

NuScale Response to NRC Request for Additional Information eRAI No. 9282 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.: 9282 Date of RAI Issue: 01/09/2018 NRC Question No.: 03.11-17 Regulatory Basis 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 within the limits of 10 CFR Part 20. 10 CFR 50.49(e)(4) requires applicants to identify the type of radiation and the total dose expected during normal operation over the installed life of the equipment. Appendix A to Part 50General Design Criteria (GDC) for Nuclear Power Plants, Criterion 61Fuel storage and handling and radioactivity control, requires systems which may contain radioactivity to be designed with suitable shielding for radiation protection and with appropriate containment, confinement, and filtering systems. GDC 4 requires applicants to ensure that structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation. 10 CFR Part 20 requires the use of engineering features to control and minimize the amount of radiation exposure to occupational workers, from both internal and external sources. NuScale DSRS 12.2 DSRS and DSRS 3.11 Acceptance Criteria states that the applicant should describe the radiation fields in sufficient detail for evaluating the inputs to shielding codes, and determination of radiation dose to electrical equipment important to safety as described in 10 CFR 50.49, and GDC 4.

Background

NuScale DCD, Tier 2 Revision 0, Table 3C-6: Normal Operating Environmental Conditions, states that the 60 Years Integrated N Dose (Rads) for the area outside of the top of the Pressurizer is 6.00E7 rads (120 rads/hour). NuScale DCD, Tier 2, Revision 0, Chapter 12.2 Table 12.2-1: Core and Coolant Source Information, only provides the fission neutron source strength and the fission neutron spectrum, without specifying the neutron energy spectrum in areas such as above the pressurizer. NuScale Technical Report TR-0116-20781-P Rev. 0 Fluence Calculation Methodology and Results, Table 5-1 Best estimate of fluence expected to be experienced in various NuScale Power Module components and locations, describes the neutron fluence to the reactor vessel and containment vessel, in the vicinity of the core, but does not provide any neutron flux and spectrum information for the area above the pressurizer.

The Control Rod Drive Mechanisms (CRDM) are located in the area above the pressurizer and inside the containment vessel. DCD Tier 2, Revision 0, Table 4.5-1, Control Rod Drive NuScale Nonproprietary

Mechanism Materials, states that Stellite 6 may be used for Hard facing for latch arm tips, and the control rod remote disconnect expansion plugs use Haynes Alloy 25. Industry material specification data shows that for Stellite 6 Haynes Alloy 25, over 50% of the base metal consist of cobalt. Industry literature also shows that Alloy X-750 (UNS N07750) the cobalt impurity is limited to 1%. The CRDM springs use springs use Alloy X-750 (UNS N07750). Due to the relatively high neutron absorption cross section of cobalt, the relatively high resultant specific radioactivity and the quantity and energy of the emitted photons when cobalt 60 (Co-60) decays, the resultant gamma dose rates, the neutron spectrum and flux where cobalt is present, are important aspects of the DSRS 12.2 and DSRS 3.11 reviews.

The neutron spectrum and flux information evaluated during the staff review under NuScale DSRS 12.2, are used in the evaluation performed by the staff for NuScale DSRS 12.3-12.4 and DSRS 3.11, related to the acceptability of the shielding design, the establishment of radiation zones, the impact on systems, structures and components, and the activation of material.

NuScale DSRS 12.2 Acceptance Criteria, states that the source descriptions should include all pertinent information required for input to shielding codes used in the design process, establishment of related facility design features, and determination of radiation dose to electrical equipment important to safety as described in 10 CFR 50.49, and GDC 4, as well as the controlling radiation exposure to workers, consistent with 10 CFR 20 and GDC 61. DSRS 12.2 also states that unless described within other sections of the FSAR, source descriptions should include the methods, models, and assumptions used as the bases for all values provided in FSAR Section 12.2. These acceptance criteria are consistent with the relevant requirements of 10 CFR Part 50 and 10 CFR Part 52.

Key Issue: The neutron flux and energy spectrum are not well-defined nor is the derivation of these values. The staff needs to know the neutron flux, energy spectrum and appropriate supporting information to evaluate the impact on materials and components located inside of the containment vessel and above the reactor core. This information is needed to evaluate the environmental qualification of components, to assess the generation of activated corrosion products, and to confirm direct occupational radiation exposure of workers during refueling evolutions. Based on information made available to the staff during the RPAC Chapter 12 Audit, the staff was not able to characterize the neutron radiation fields in the aforementioned areas.

Question To facilitate staff understanding of the application information sufficient to make appropriate regulatory conclusions, the staff requests that the applicant:

Identify and describe the methods, models and assumptions used to calculate the neutron spectrum and flux above the top of the pressurizer, inside the containment vessel.

Provide data in NuScale DCD, Tier 2, Revision 0, Section 12.2 describing the neutron spectra and flux, at the area identified above, and the assumptions and input parameters used.

NuScale Nonproprietary

OR Provide the specific alternative approaches used and the associated justification.

NuScale Response:

NuScale modeled the core neutron source term (as provided in FSAR Table 12.2-1) using a U-235 Watt fission spectrum with a NuScale specific neutron source strength intensity. The maximum neutron release rate for the modeled core design configurations (U-235 enrichment and Gd2O3 loading) and burnup was used to conservatively adjust the neutron source term, thus creating a bounding neutron source term. In addition, this conservative neutron source term is multiplied by an assembly peaking factor of 1.461. The MCNP6 code is then used to transport the core neutron source term through the modeled NPM to determine the fluxes and dose rates in the upper parts of the module, such as at the top of the pressurizer inside the containment vessel. Neutron flux to dose conversion factors from ANSI/ANS-6.1.1 was used in MCNP to calculate the dose rates from neutron flux. The neutron energy spectra used for component activation in the core was the previously described fission neutron energy distribution, (i.e. a Watt Fission spectrum for U235). For the sources generated by neutron activation, the associated ex-core neutron flux is described in associated FSAR Tables 12.2-23 and 12.2-25.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary