ML18092B090

University of Massachusetts Lowell - Request for Additional Information Regarding the Renewal of Facility Operating License No. R-125 for the University of Massachusetts Lowell Research Reactor and Primarily Related to Instrumentation and C

ML18092B090
Person / Time
Site:	University of Lowell
Issue date:	07/19/2019
From:	Duane Hardesty, Edward Helvenston Office of Nuclear Reactor Regulation
To:	Prosanta Chowdhury Univ of Massachusetts - Lowell
Helvenston E, NRR/DLP 415-4067
References
EPID L-2015-RNW-0001
Download: ML18092B090 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 19, 2019 Dr. Partha Chowdhury, Director Nuclear Radiation Laboratory University of Massachusetts-Lowell One University Avenue Lowell, MA 01854

SUBJECT:

UNIVERSITY OF MASSACHUSETTS LOWELL RESEARCH REACTOR-REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-125 FOR THE UNIVERSITY OF MASSACHUSETTS LOWELL AND PRIMARILY RELATED TO INSTRUMENTATION AND CONTROLS (EPID NO. L-2015-RNW-0001)

Dear Dr. Chowdhury:

The U.S. Nuclear Regulatory Commission (NRC) staff is continuing its review of the University of Massachusetts Lowell (UML)s application for the renewal of Facility Operating License No. R-125 for the UML Research Reactor dated October 20, 2015 (Agencywide Documents Access and Management System Accession No. ML16042A015), as supplemented.

The NRC staff identified additional information needed to continue its review of the renewal request, as described in the enclosed request for additional information (RAI). As discussed by telephone on July 16, 2019, provide a response to the RAI, or a written request for additional time to respond, including the proposed response date and a brief explanation of the reason, by October 21, 2019. Following receipt of the complete response to the RAI, the NRC staff will continue its review of the renewal request.

The response to the RAI must be submitted in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.4, Written communications, and, per 10 CFR 50.30(b), Oath or affirmation, be executed in a signed original document under oath or affirmation. Information included in the response that you consider sensitive or proprietary, and seek to have withheld from public disclosure, must be marked in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding. Any information related to safeguards should be submitted in accordance with 10 CFR 73.21, Protection of Safeguards Information:

Performance Requirements.

Based on the response date provided above, the NRC staff expects to complete its review and make a final determination on the renewal request by June 30, 2020. This date could change due to several factors including a need for further RAIs, unanticipated changes to the scope of the review, unsolicited supplements to the application for renewal, and others. If the forecasted date changes, the NRC staff will notify you in writing of the new date and an explanation of the reason for the change. In the case that the NRC staff requires additional information beyond that provided in the response to this RAI, the NRC staff will request that information by separate correspondence.

If you have any questions regarding the NRC staffs review or if you intend to request additional time to respond, please contact me at 301-415-4067 or by electronic mail at Edward.Helvenston@nrc.gov.

Sincerely,

/RA/

Edward Helvenston, Project Manager Research and Test Reactors Licensing Branch Division of Licensing Projects Office of Nuclear Reactor Regulation Docket No. 50-223 License No. R-125

Enclosure:

As stated cc: See next page

University of Massachusetts - Lowell Docket No. 50-223 cc:

Mayor of Lowell City Hall Lowell, MA 01852 Mr. Leo Bobek Reactor Supervisor University of Massachusetts - Lowell One University Avenue Lowell, MA 01854 Department of Environmental Protection One Winter Street Boston, MA 02108 Jack Priest, Director Radiation Control Program Department of Public Health Schrafft Center, Suite 1M2A 529 Main Street Charlestown, MA 02129 John Giarrusso, Planning and Preparedness Division Chief Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 Test, Research and Training Reactor Newsletter Attention: Ms. Amber Johnson Department of Materials Science and Engineering University of Maryland 4418 Stadium Drive College Park, MD 20742-2115

ML18092B090 *concurred via email NRR-088 OFFICE NRR/DLP/PRLB/TR*

NRR/DLP/PRLB/PM*

NRR/DLP/PROB/LA*

NAME DHardesty EHelvenston NParker DATE 07/16/2019 07/17/2019 07/16/2019 OFFICE NRR/DPR/PRLB/BC NRR/DPR/PRLB/PM NAME GCasto EHelvenston DATE 07/19/2019 07/19/2019

Enclosure OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RENEWAL OF THE UNIVERSITY OF MASSACHUSETTS LOWELL RESEARCH REACTOR LICENSE NO. R-125; DOCKET NO. 50-223 The U.S. Nuclear Regulatory Commission (NRC) is continuing its review of the University of Massachusetts Lowells (UMLs) application for the renewal of Facility Operating License No. R-125, for the UML Research Reactor (UMLRR), dated October 20, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16042A015), as supplemented.

The NRC staff conducted an onsite regulatory audit (Audit) to review the UMLRR instrumentation and control (I&C) upgrade on July 26-28, 2017, in accordance with the regulatory audit plan (ADAMS Accession No. ML17220A243). The purpose of the Audit was to determine if the design and development processes used, and the outputs of those processes, resulted in upgrades to I&C systems that meet applicable regulatory requirements. In addition, the Audit was conducted to address open-items and identify information that would be required to be docketed in order to support the basis of the licensing decision and allow the NRC staff to perform a more efficient review of the UMLRR I&C upgrades.

The NRC staff identified several requests for additional information (RAI), which were discussed during the Audit and are necessary to support NRC review. We request that you provide responses to the following RAIs by October 21, 2019. UML may provide alternative justification that demonstrates the ability of UMLRR to maintain and perform the safety function(s) associated with these RAIs.

The license renewal application is evaluated using the appropriate regulations in Title 10 of the Code of Federal Regulations (10 CFR), and the following guidance:

NUREG-1537, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content, issued February 1996 (ADAMS Accession No. ML042430055)

NUREG-1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, issued February 1996 (ADAMS Accession No. ML042430048)

American National Standards Institute/American Nuclear Society, (ANSI/ANS) 15.1-2007 (R2013), The Development of Technical Specifications for Research Reactors.

RAI-7.14 and RAI-7.16 below are related to the proposed technical specifications (TSs) provided on March 5, 2019, as a supplement to UMLs renewal application (ADAMS Accession No. ML19064B377). TSs are fundamental criteria necessary to demonstrate facility safety and are required by 10 CFR 50.36, Technical specifications, for each license authorizing operation of a production or utilization facility of a type described in 10 CFR 50.21, Class 104 licenses; for medical therapy and research and development facilities. TSs are derived from the analyses and evaluation included in the safety analysis report (SAR) and submitted pursuant to 10 CFR 50.34, Contents of applications; technical information. TSs for nuclear reactors will include items in the following categories: safety limits (SLs), limiting safety system settings (LSSSs), limiting conditions for operation (LCO), surveillance requirements, design features, and administrative controls. The NRC guidance for TSs is provided in NUREG-1537, Part 1, Appendix 14.1, Format and Content of Technical Specifications for Non-Power Reactors. This guidance relies significantly on ANSI/ANS-15.1-2007. The NRC staff takes the position that the statements in these documents provide acceptable guidance to licensees and, unless acceptable alternatives are justified by the licensee, should be utilized whenever appropriate.

RAI-7.1 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the structures, systems, and components (SSCs) of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Additionally, the regulations in 10 CFR 50.34(b)(6)(iv) state, in part, that final safety analysis report shall include plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of SSCs.

The guidance in NUREG-1537, Part 1, Section 7.1, states, in part, the applicant should describe the instrumentation and control (I&C) systems, including block, logic, and process flow diagrams showing major components and subsystems, and connections among them. Additionally, the applicant should summarize the technical aspects, safety, philosophy, and objectives of the I&C system design recommended by Section 7.2 and should discuss such factors as redundancy, diversity, and isolation of functions.

Provide the following, or justify why no additional information is required:

a. Details are provided on the I&C systems in the UMLRR SAR, predominantly in Chapter 7. However, it is not clear which I&C systems and descriptions existed at the time Facility Operating License No. R-125 Facility License was last renewed (or those I&C systems that were the subject of a license amendment related to I&C systems since the last renewal); which I&C systems were implemented without prior NRC approval since the last renewal; and which I&C systems are proposed as part of the current license renewal application.

Provide a detailed description of the changes proposed in license renewal for the I&C systems that includes the recommended details of Sections 3.1, 7.1, and 7.2 of NUREG-1537, Part 1. The description should reference the specific SAR sections (or SAR supplements provided in support of license renewal) that discuss the systems that are the subject of the proposed changes. The description should:

describe the design of the proposed Thermo Scientific Wide Range Log Period Power Module (PPM) and the General Atomics (GA) NMP-1000 Modules and discuss how they meet the UMLRR design criteria of the flux monitoring systems, such as:

design for the complete range of normal, transient, and accident conditions, quality standards commensurate with the safety function they provide; and, the reactor protection system (RPS) design criteria identified in Section 7.2.1 of the UMLRR SAR.

include a detailed discussion of any differences (e.g., scram detector voltage, etc.)

between the proposed systems and any existing systems that are being removed or replaced, discuss any proposed TSs, including surveillance TSs and intervals, and the detailed design bases for the TS specific to design and operation that are necessary to ensure the availability and operability of the proposed I&C systems (e.g., Thermo Scientific Wide Range Log PPM and GA NMP-1000s).

b. The UMLRR SAR describes many alarm and trip functions that activate trip relays or contacts in the RPS scram circuit. However, the SAR does not appear to provide a diagram for the scram circuit train showing the arrangement and configuration of the circuit.

Provide a diagram depicting the overall trip circuit showing how each circuit is arranged to ensure a protective system action interrupting the scram circuit.

RAI-7.2 The regulations in 10 CFR 50.59, Changes, test and experiments, state, in part, that a licensee may make changes to the facility and procedures as described in the SAR if the licensee makes a determination (documented in a written evaluation) that no TS change is required, and that the changes do not meet any of the criteria in 10 CFR 50.59(c)(2).

Section 50.34(a)(7) of 10 CFR requires an applicant to describe the quality assurance (QA) program for design, fabrication, construction, and testing of the SSCs of the facility, and 10 CFR 50.34(b)(6)(ii) requires that a final safety analysis report include the managerial and administrative controls to be used to assure safe operation. Section 50.34(b)(2) of 10 CFR requires a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.

Section 7.6.1.1 of the UMLRR SAR states that the human machine interface (HMI) display screens use control and input/output hardware manufactured for industrial process control, process monitoring, and data acquisition. The software used to operate the system is stated to be an integrated suite of industrial control and automation software provided by the hardware manufacturer. Section 7.6.1.1 of the SAR also states that the ARMS [area radiation monitoring system] was installed in 1999, the PCS [process controls and instrumentation system] in 2001, and the DCS [display control system] in 2003 under 10 CFR 50.59. SAR Section 7.6.1.3 states that each HMI computer (i.e., PCS, DCS, and ARMS) displays the information from the associated controller and provides a touch screen and a keyboard terminal for manual control of process functions. The SAR further states that, [t]he display configurations were developed using the integrated software package associated with the hardware.

Consistent with the regulations in 10 CFR 50.34, provide the following information for the HMI displays and integrated software, which does not appear to be included in the SAR, as supplemented, or discuss why no additional information is required:

a. Describe the development of the display configurations, including the name of the integrated software package and verification that a structured process was used in developing the software for both safety and non-safety-related systems.

b. Explain how functional characteristics of the software requirements specifications were properly (and precisely) described for each requirement.

c. Describe how the software and hardware were integrated; and discuss how UML verifies the required computer system software is installed in the appropriate system configuration, and how UML ensures that the correct version of the software/firmware is installed in the correct hardware components, to verify operability.

d. Identify the developer of the integrated systems software/hardware for the PCS, DCS, and ARMS and describe the verification and validation (V&V) or other testing performed to assure software operability and reliability.

RAI-7.3 The regulations in 10 CFR 50.59 state, in part, that a licensee may make changes to the facility and procedures as described in the SAR if the licensee makes a determination (documented in a written evaluation) that no TS change is required, and that the changes do not meet any of the criteria in 10 CFR 50.59(c)(2). Section 50.34(b)(4) of 10 CFR requires a final analysis and evaluation of the design and performance of SSCs and taking into account any pertinent information developed since the submittal of the preliminary safety analysis report.

Table 1-5 of the UMLRR SAR lists facility modifications since the last renewal for which UML states a 10 CFR 50.59 review was performed to ensure that a license amendment was not required. However, the NRC staff notes that Table 1-5 does not appear to include some 10 CFR 50.59 reviews that are described elsewhere in the SAR or in annual reports.

Specifically, SAR Section 7.6.1.1 states that the ARMS was installed in 1999 under 10 CFR 50.59. Additionally, the 2009-2010 annual report (ADAMS Accession No. ML102460026) and the 2015-2016 annual report (ADAMS Accession No. ML16224A324) for the UMLRR describe changes related to 10 CFR 50.59, including drives control system, and control room annunciator panel replacement, respectively (the NRC staff notes that the annunciator panel replacement may have been reviewed/implemented after the 2015 SAR was submitted).

Additionally, it is not clear to the NRC staff that all of the 10 CFR 50.59 reviews that appear to be relevant to the current UMLRR I&C system, are fully described in the SAR, as supplemented.

For example, upgrades to drive position are described in the SAR, but these do not appear to be the same changes that are indicated in the 2003, Drives Control System, change listed in SAR Table 1-5.

Provide the following, or justify why no additional information is required:

a. Clarify whether SAR Table 1-5 provides a complete list of 10 CFR 50.59 reviews, and if necessary, provide a revised and updated list of 10 CFR 50.59 reviews since the previous license renewal, which includes items discussed elsewhere in the SAR and/or in annual reports, as appropriate. Also, clarify whether the items in the list received a full 10 CFR 50.59 review and written evaluation, or whether some items in the list were reviewed for 10 CFR 50.59 applicability, but were determined to screen out of needing a full review (based on the NRC staffs review of UMLRR annual reports, it appears that UML may have determined that some of the items listed in SAR Table 1-5 screened out). Additionally, clarify whether the items in the list have all been implemented, or whether the list includes items that were reviewed but not implemented (e.g., SAR Table 1-5 includes a Linear Channel Replacement, in 2014, but it is not clear whether this is the channel replacement that UML is requesting that the NRC review in conjunction with its renewal request, or a different change that has already been implemented).

b. Provide copies of the 10 CFR 50.59 reviews (or screens) listed in RAI Table 7-1 below, to allow the NRC staff to effectively evaluate the current I&C system for license renewal.

Additionally, provide copies of the 10 CFR 50.59 reviews (or screens) for any other modifications to the facility since the last license renewal that UML determines could impact the current (or proposed) UMLRR I&C systems, if any. Alternatively, provide a detailed evaluation and discussion of the changes listed below (and any other relevant changes), including descriptions of any modification or addition to, or removal from, the facility or procedures that affects the I&C system design function, or method of performing or controlling the function; and, their impact on UMLRR operations.

c. During the Audit, the NRC staff reviewed a 10 CFR 50.59 screen numbered 16-01 related to the addition of control room alarms and indicators. Clarify whether this screen is the same as the Control Room Annunciator Panel Replacement (discussed in the 2015-2016 annual report) item listed in RAI Table 7-1 below. If 16-01 is a different screen, provide the information requested in RAI-7.3 a. and b. above.

RAI Table 7-1: UMLRR 10 CFR 50.59 Reviews (or Screens)

Year Title Reviewed 2016 Control Room Annunciator Panel Replacement*

?

2014 Linear Channel Replacement 12/17/2014 2014 Addition of Panel Indicators 10/7/2014 2013 Log-N Channel Replacement 12/18/2013 2012 Chart Recorder Replacement 01/12/2012 2010 Drives Control System*

?

2008 Secondary Cooling System Remote Control 03/14/2008 2003 Drives Control System 02/20/2003 2001 Upgrade of UMLRR Process Control Cabinet 10/25/2001 1999 ARMS Upgrade**

?

1998 Power Detector Mechanical Height Adjusters 12/17/1998 1997 Instrumentation Upgrade 09/11/1997 Estimated date (actual date unknown)

Discussed in UMLRR Annual Reports

• • Discussed in Section 7.6.1.1 of the UMLRR SAR RAI-7.4 The regulations in 10 CFR 50.9, Completeness and accuracy of information, require that information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects. The regulations in 10 CFR 50.30(b) Oath or affirmation, require that license applications, and each amendment to such applications, must be executed in a signed original by the applicant or duly authorized officer thereof under oath or affirmation.

By letter dated July 26, 2017 (ADAMS Accession No. ML17249A080), GA submitted, to the NRC, a request for withholding of information from public disclosure under 10 CFR 2.390, Public inspections, exemptions, requests for withholding, for documents related to its NMP-1000 Multi-Range Linear Module. GAs letter stated that these documents were being submitted in support of licensing requests, planned or pending, with the NRC initiated by UML.

The request for withholding was granted to GA by letter dated December 7, 2018 (ADAMS Accession No. ML18338A040). However, the NRC requires that all design documentation to be considered by the NRC staff in the licensing determination for the UMLRR I&C upgrade be on the UML docket.

Provide the following, or justify why no additional information is required:

a. During the Audit, the NRC staff reviewed available documentation regarding the design criteria and bases for the GA NMP, NLW, and NLX systems; however, this documentation is proprietary, and this GA documentation indicates that it is documentation prepared for another facility. Indicate which, if any, of the below documents are applicable, and the extent of their applicability to the UMLRR I&C install:

1)

T3322000-1AT_Rev A NLW-1000 Acceptance Test Procedure

2)

T3322000-RTM_Rev A NLW-1000 Traceability Matrix - easy read

3)

T3322000-RTM_Rev A NLW-1000 Traceability Matrix

4)

T3401000-1AT_Rev A NMP-1000 Acceptance Test Procedure

5)

T3401000-RTM_Rev A NMP-1000 Traceability Matrix - easy read

6)

T3401000-RTM_Rev A NMP-1000 Traceability Matrix

7)

T9S900D940-SYR_Rev A NMP-1000 System Requirements Specification (SyRS)

8)

T9S900D941-SRS, NMP-1000 Software Requirements Specification (SRS)

9)

T9S900D950-SYR_Rev A NLX-1000 SyRS

10) T9S900D951-SRS, NLX-1000 SRS

11) T9S900D980-FME Rev A, NMP-1000 Failure Modes Effects Analysis

12) 20130207001-RPT Rev 2, NetBurner-MOD54415 Validation Summary Report

13) LPC E117-1017 Rev 1, NMP1000 operations and maintenance manual

14) T3401000-1UMB, NMP-1000 user manual

b. Additionally, to meet the requirements of 10 CFR 50.9, 10 CFR 50.30, and 10 CFR 2.390, submit, under oath or affirmation, a request for the above documentation (listed in RAI-7.4.a) to be placed, as applicable, on the 50-223 docket with reference to the affidavit from the content owner justifying withholding stating the above documentation applies to the renewal of Facility Operating License No. R-125.

c. During the Audit, the NRC staff identified additional information that is necessary for the NRC staff to reach a licensing decision. Some of the documentation listed above (e.g., item 13 and 14) may be necessary for the NRC review, but was not submitted under the GA affidavit dated July 26, 2017. Submit a copy of that documentation under oath or affirmation, along with a 10 CFR 2.390 affidavit signed by the content owner (if appropriate).

d. During the Audit and in a supplement to its license renewal application (ADAMS Accession No. ML19100A273), UML proposed a replacement for the Log Power Measuring Channel, which uses the Thermo Scientific PPM to replace the existing GA wide-range logarithmic power module. Provide similar documentation for the proposed PPM commensurate with the GA documentation referenced in 7-4.a above (e.g., Thermo Scientific Ex-core Neutron Flux Monitoring System Instruction Manual: 937 Wide Range Signal Processor Document number 937 Revision, and similar design, operations, and maintenance documentation).

RAI-7.5 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 7.4, states that, [h]ardware and software for computerized systems should meet the guidelines of [Institute of Electrical and Electronics Engineers] (IEEE) 7-4.3.2-1993. IEEE Standard 7-4.3.2-2010, Clauses 5.3, 5.3.1, 5.3.5, and 5.3.4, recommend that the development and the integration of computer hardware and software be addressed in the development process, including software QA and the use of software quality metrics to assess whether software quality requirements are being met throughout the system lifecycle, V&V activities, and configuration management to ensure changes to the software/firmware are formally documented and approved.

Section 1.4.3 of the NMP-1000 Software Requirements Specification, T9S900D941-SRS, references several GA-EIS documents including the software QA plan, software V&V plan, and other documents. Section 3.6.4.1 of the NMP-1000 System Requirements Specification, T9S900D940-SYR (item No. 7 listed in RAI-7.4.a above), states that, [s]oftware configuration shall be formally controlled according to the Software Configuration Management Plan.

However, the details of these plans for the NMP-1000 are not clear from the SAR or other information submitted in support of license renewal. Additionally, similar design-specific documentation does not appear to have been provided for other I&C upgrades proposed in conjunction with license renewal (e.g., Thermo Scientific Wide Range Log PPM).

Provide the following, or justify why no additional information is required:

a. Provide the specific vendor plans (e.g., those listed above), or reference to the vendors process certifications to ensure conformance to industry standards on software QA (such as the International Organization for Standardization (ISO) 9000 or a process level improvement model, such as Capability Maturity Model Integration (CMMI)), that document the formal process for the NMP-1000 upgrades that describes the software development through the system lifecycle to ensure software quality and configuration management. Alternatively, provide equivalent details on UML processes and plans that similarly demonstrate software quality and configuration management.

b. Provide similar design-specific documentation for all other proposed I&C upgrades (e.g.,

Thermo Scientific Wide Range Log PPM). Appropriate documentation may be of the types listed below. (Note: the below list is typical vendor information of the type that would aid in the NRC staff review. However, these are examples and not necessarily indicative of actual required documents.)

Software Development Plan (SDP)

Software Configuration Management Plan (SCMP)

Software Quality Assurance Plan (SQAP)

Software Integration Plan (SIntP)

Software Installation Plan (SInstP)

Software Maintenance Plan (SMaintP)

Software Safety Plan (SSP)

Software Verification and Validation Plan (SVVP)

Software Configuration Management Plan (SCMP)

Software Test Plan (STP)

RAI-7.6 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 4, states that ANSI/ANS 15.20 is useful as general guidance for the design, implementation, and evaluation of I&C systems. Clauses 9.3, 9.3.1, and 9.3.2 of ANSI/ANS 15.20 state that verification is used during software development to facilitate elimination of errors in computer-based systems, and that a key ingredient in the validation process is sufficient independence of the review team. NUREG-1537, Part 2, Section 7.4, states that, [h]ardware and software for computerized systems should meet the guidelines of IEEE 7-4.3.2-1993. Clauses 5.3.3 and 5.3.4 of IEEE Standard 7-4.3.2-2010 state that the V&V process applies to both hardware and software. However, the details of V&V for the proposed I&C upgrades are not clear to the NRC staff from the SAR or other information submitted in support of license renewal.

Describe any V&V performed for the NMP-1000 modules, Thermo Scientific PPM, and other proposed I&C upgrades. Additionally, provide documentation that shows the plan for the V&V and V&V procedures, and when available, the results confirming V&V has been successfully accomplished. These documents should demonstrate that the proposed I&C systems meet the design requirements and specifications, and that it fulfills its intended purpose. The V&V testing should also show that the test plan and test results, including any deviations, required corrective actions, and retests were reviewed by the Reactor Safety Subcommittee (RSSC) and reviewed and approved by UMLRR Reactor Supervisor. Alternatively, discuss why no additional information is required.

RAI-7.7 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 7.4, states that, [h]ardware and software for computerized systems should meet the guidelines of IEEE 7-4.3.2-1993. IEEE Standard 7-4.3.2-2010, Clause 5.3.5, states that, [c]hanges to the software shall be formally documented and approved consistent with the software configuration management plan.

Software configuration management (CM) should include a determination that any software modifications, including firmware, during the design process, and after acceptance of the software for use, will be made to the appropriate version and revision of the software. However, the details of software CM for the proposed I&C upgrades are not clear from the SAR or other information submitted in support of license renewal.

Describe how software changes after initial delivery will be reviewed, tracked and documented.

Additionally, provide documentation that a configuration management program appropriately traces changes to safety system softwarefrom their point of origin to implementationand addresses any impacts on system safety, control console, or display instruments. Alternately, explain why configuration control is not required, or discuss why no additional information is required.

RAI-7.8 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 7.4, states that, [h]ardware and software for computerized systems should meet the guidelines of IEEE 7-4.3.2-1993. IEEE Standard 7-4.3.2-2010, Clause 5.11, states that, [s]oftware and hardware identification, including version control, shall be provided and used to verify that the correct software is installed in the correct hardware component. The digital computer system equipment for the displays and processorincluding hardware, software, firmware, and interfacesshould be reviewed to provide assurance that the required computer system hardware and software are installed on the appropriate system configuration. However, the details of how this will be done for the proposed I&C upgrades does not appear to be provided in the SAR or other information submitted in support of license renewal.

a. Provide a description of any applicable program or procedure used to ensure that the correct version of the software is installed on the NMP-1000 modules and demonstrate assurance that the required computer system hardware and software are installed in the appropriate system configuration, including a program to ensure that the correct version of the software/firmware is installed in the correct hardware components (i.e., as part of the reactor startup checklist procedure to verify operability).

b. During the Audit, UML staff confirmed the NRC staff observation that the NMP-1000 user manual stated that GA would do all software installs. Explain how UML will verify the version of the software install is correct, and how UML will validate that any updates are correct for the UMLRR, including any updates to the programs and procedures addressed in RAI 7.8.a.

RAI-7.9 The regulation in 10 CFR 50.34(b)(6)(iv) states, in part, that the final safety analysis report shall include plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of SSCs. The regulation in 10 CFR 73.40, Physical protection: General requirements at fixed sites, requires that licensees provide physical protection at a fixed site, or contiguous sites where licensed activities are conducted, against radiological sabotage, or against theft of special nuclear material, or against both.

The guidance in NUREG-1537, Part 2, Section 4, states that the guidance in ANSI/ANS-15.15-1978 is useful as general guidance for the design, implementation, and evaluation of I&C systems. The guidance in ANSI/ANS-15.15-1978, Clause 5.10, states that physical provisions shall be provided to prevent the unauthorized use of the reactor controls and limit access to setpoint and calibration adjustments to the extent necessary to prevent inadvertent misadjustments. Access control includes cyber security vulnerabilities (physical and electronic), including preventing/limiting unauthorized physical and electronic access during the developmental or operational phase and the transition from development to operations.

During the Audit, the NRC staff reviewed the NMP-1000 user manual and observed that, unless disabled or administratively controlled, the NMP-1000 modules can accept commands via the Ethernet port (J9 connector) or the analog remote interface connector on the rear panel (J8 connector). The J8 connection uses an RS-232 cable remote display and a null-modem cable connected to a PC for maintenance.

a. During the Audit, UML staff indicated that the Ethernet connection (J9) will not be used.

However, when the UML staff removed the NMP-1000 outer case, the NRC staff observed that the ethernet capabilities were not disabled (i.e., the cord was still attached). Verify that the maintenance connection (J8) and the ethernet connection (J9) will not be used, that the internal cable connector for J9 was removed, and state how this configuration will be assured.

b. During the Audit, UML staff stated that GA will perform any and all maintenance on the NMP-1000s. Explain how UML staff will ensure configuration management will be maintained, and how necessary post-maintenance testing will be performed.

Additionally, discuss the UML staffs role in verifying and approving the test results are acceptable prior to resuming operations with a channel on which GA has performed maintenance to ensure TS-required operability.

c. During the Audit, UML staff also indicted that the remote interface push button is for use in a GA console, which the UMLRR does not have. The UML staff stated that pushing the remote interface push button disables the ability to reset a scram. Explain how UML will ensure that the remote input to the NMP-1000 modules will not be used to provide commands to the NMP-1000 modules.

RAI-7.10 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent. The regulations in 10 CFR 50.34(b)(6)(iv) state, in part, that the final safety analysis report shall include plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of SSCs. Additionally, the regulations in 10 CFR 50.9 require that information provided to the Commission by an applicant for a license or by a licensee, or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee, shall be complete and accurate in all material respects.

Provide the following, or justify why no additional information is required:

a. During the Audit, the NRC staff reviewed UML 10 CFR 50.59 screen 16-01 related to the addition of control room alarms and indicators, and also observed that the additional alarms and indicators were added in the control room (the additional alarms and indicators are on a new/replaced control room annunciator panel). The control room instrument panel, which is separate from the control console, houses the control room annunciator panel. The NRC staff observed that the new annunciator panel is a 5 x 5 panel (i.e., laid out with 5 rows of indicator spaces with 5 indicator spaces each).

The NRC staff also reviewed the UMLRR 2015-2016 annual report, which describes an annunciator panel upgrade performed by UML without prior NRC review or approval (see also RAI-7.3). The annual report states that the original panel had 17 individual alarm indicators (2 by 3 inches each), while the new panel has 23 indicators (2.25 by 2.75 inches each); 6 additional alarm indicators were added to provide additional information for the operator. The annual report also states that the new alarm panel performs the same function as the old panel, but the new panel uses more reliable technology (light-emitting diodes rather than incandescent lights). The annual report states that [t]he indicators are slightly larger and have the same physical location in the control room.

The control room annunciator panel is labelled P-22 in SAR Figure 7-8, as well as in updated Figure 7-9 in UMLs supplemental information dated April 10, 2019 (ADAMS Accession No. ML19100A273). SAR Figure 7-8 shows a 4 x 4 panel. SAR Section 7.4.4 lists 16 alarm conditions, and SAR Table 7-7 states that the alarm panel

[p]rovides annunciator buzzer and annunciator lights for 16 monitored conditions.

Updated SAR Figure 7-9 in UMLs supplemental information dated April 10, 2019, also shows a 4 x 4 panel.

i.

Given that the SAR, as supplemented, appears to the NRC staff to describe the old control room annunciator panel instead of the new panel, provide updated SAR section(s) or detailed descriptions reflecting the new 5 x 5 panel and the additional information provided to the operator that was not available with the old 4 x 4 panel.

Include a discussion of the additional alarm indicators included on the new panel.

ii. The NRC staff noted that there appear to be discrepancies in the descriptions of the old and new annunciator panels based on the information in the SAR, as supplemented; information in the 2015-2016 annual report; and the NRC staffs observations during the Audit. Explain why the SAR, as supplemented, appears to state that the old panel monitored 16 alarm conditions, but the annual report appears to state that the old panel monitored 17 conditions. Clarify why the new panel only has 23 indicators, if it is a 5 x 5 panel with 25 indicator spaces (e.g., because 2 spaces are unused, if this is the case). Also, explain the statement in the annual report that the new indicators are slightly larger, given that the dimensions provided for the new indicators appear to be smaller. Additionally, if there are any discrepancies or differences between the 10 CFR 50.59 screen 16-01 (requested in RAI-7.3) and the annual reports or the SAR, as supplemented, explain these discrepancies.

b. The guidance in NUREG-1537, Part 2, Section 7.4, states that, [h]ardware and software for computerized systems should follow the guidelines of IEEE 7-4.3.2-1993. IEEE Standard 7-4.3.2 provides guidance to address system design risks created by human errors in the operation and support of the system. To determine if the human-system interface (HSI) aspects of a display modification have an adverse effect on SAR-described design functions, potential impacts due to the number and/or type of parameters displayed by and/or available from the HSI should be evaluated.

Consideration of a digital modification's impact due to the number and/or type of parameters displayed by and/or available from the HSI should involve an examination of the actual number and/or type of parameters displayed by and/or available from the HSI and how they could impact the performance and/or satisfaction of SAR-described design functions. An increase in the amount of information that is provided such that the amount of available information could have a detrimental impact on the operator's ability to discern a particular condition or to perform a specific task. The evaluation should also consider logical grouping and relevance.

Provide an evaluation that considers the HSI impact of the additional indicators, and that addresses potential hazards created by human errors in the operation of the system to assure that the functions allocated in whole or in part to the human operator(s) and maintainer(s) can be successfully accomplished to meet the safety system design goals.

Include an explanation of any changes made in procedures or operating instructions to mitigate the potential adverse impact.

RAI-7.11 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 7.4, states that, [t]he scram system should be designed to annunciate the channel initiating the action, and to require a resetting to resume operation. NUREG-1537, Part 2, Section 4, states that the guidance in ANSI/ANS 15.15-1978 is useful as general guidance for the design, implementation, and evaluation of I&C systems.

The guidance in ANSI/ANS-15.15-1978, Clause 5.8, states that, [e]ach channel shall indicate in a distinctive manner when it is in the tripped state, and that [o]nce tripped, the RSS [reactor safety system] shall remain in the tripped state at the system level and shall indicate the protective instrument subsystem initiating the shutdown until deliberate action is taken by the reactor operator.

During the Audit, the NRC staff viewed the new indicator and alarm panel, the PCS display, and the DCS display (P22, C10, and C13, respectively, in updated Figures 7-9 and 7-10 in UMLs supplemental information dated April 10, 2019), which indicate the cause of a UMLRR alarm.

As described in Chapter 7 of the SAR, the RSS channels are designed with 1/N logic; any one of N signal inputs to either logic unit will cause the trip actuator amplifiers to trip and initiate a reactor scram. During the Audit, the NRC staff reviewed the NMP-1000 user manual and observed that the NMP-1000 modules have a trip reset function on the main menu. However, it is not entirely clear how operators can distinguish which channel causes a trip, or if all channels need to be reset following a trip.

Describe the information available to the operators to distinguish the channel that caused the trip and the actual cause of the trip. Additionally, explain if all channels need to be individually reset or just the channel that tripped. Alternatively, discuss why no additional information is required.

RAI-7.12 The regulations in 10 CFR 50.34(a)(7) and 10 CFR 50.34(b)(6)(ii) require that the SAR include

[a] description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the SSCs of the facility, and [m]anagerial and administrative controls to be used to assure safe operation.

The guidance in ANSI/ANS-15.8-1995 (endorsed by NRC Regulatory Guide 2.5), Clause 2.3.3, states that, [t]he need for or the use of qualification tests shall be defined in a formal test plan that shall include appropriate acceptance criteria and shall demonstrate the adequacy of performance under conditions that simulate the most adverse design conditions. Test results shall be documented and evaluated by the responsible design organization to assure that test requirements have been met.

The guidance in NUREG-1537, Part 1, Subsection 7.2.1, recommends that, [a]ll systems and components of the I&C systems should be designed, constructed, and tested to quality standards commensurate with the safety importance of the functions to be performed.

The design acceptance criteria in NUREG-1537, Part 2, Section 7.4, states that, [t]he design reasonably ensures that the design bases can be achieved, [and] the system will be built of high-quality components.

During the Audit, UML staff indicated that they performed tests of the NMP-1000 upon receipt; however, a test plan was not developed, and test results were not recorded. UML stated that a test plan and documented test results (i.e., a site acceptance test) will be developed.

Provide the following, or discuss why no additional information is required:

a. Describe how the quality of the components and modules of the I&C system upgrades (e.g., NMP-1000 and PPM) was verified to be commensurate with its safety importance.

Additionally, describe how the QA program at the UMLRR provides controls over the design, fabrication, installation, and modification of the RPS to the extent that these impact UMLRR safety-related items.

b. Describe the site acceptance testing (or other applicable V&V testing) that was performed for the NMP-1000 modules and how this testing validates that the NMP-1000s will meet the quality and design requirements for fulfilling its intended safety function as documented in Section 7.2 of the UMLRR SAR.

c. Describe the site acceptance testing (or other applicable V&V testing) that was performed for the Thermo Scientific PPM and how this testing validates that the PPM will meet the quality and design requirements for fulfilling its intended safety function as documented in Section 7.2 of the UMLRR SAR.

RAI-7.13 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

During the Audit, the NRC staff reviewed T9S900D980-FME Rev A NMP-1000 Failure Modes Effects Analysis (document No. 11 listed in RAI-7.4.a above). The NRC staff noted that Subsection 4.2.1.2 of the analysis documented that unused code was detected and subsequently removed from the NMP-1000 software. However, it is not clear if unused code was removed from the channels that are proposed to be used at the UMLRR. Additionally, the NRC staff reviewed the 20130207001-RPT, Revision 2, NetBurner-MOD54415 Module Validation Summary Report (document No. 12 listed in RAI-7.4.a above) and noted that the code review checklist documents many instances of additions and removal of software code.

These changes include bug fixes, added NMP-1000 functionality, and removal of unused code.

a. Verify and document that the unused code cited in the Failure Modes Effects Analysis was removed from subsequent NMP-1000 units, specifically, those sold to UML for installation as UMLRR equipment. Alternatively, justify why no additional information is required.

b. Verify and document that the modifications to the software code detailed in the NetBurner-MOD54415 Module Validation Summary Report were also completed for NMP-1000 units sold to UML for installation as UMLRR equipment. Alternatively, justify why no additional information is required.

RAI-7.14 The guidance in Section 7.4 of NUREG-1537, Part 2, recommends that TSs, including surveillance tests and intervals, should be based on discussions and analyses of required safety functions in the SAR and that the surveillance tests and intervals give confidence that the equipment will reliably perform its safety function. Additionally, ANSI/ANS-15.15-1978 recommends the system design include capability for periodic checks, tests and calibrations.

ANSI/ANS-15.15-1978 also recommends that, if on-line periodic testing is necessary, such testing should not reduce the capability of the system to perform its safety function.

Provide the following, or justify why no additional information is required:

a. The proposed UMLRR TSs submitted March 5, 2019, include the following:

TS 4.1(7) states, [t]he linear and logarithmic power channels signals shall be checked against a heat balance annually; TS 4.2.3(3) states, [a] channel calibration of the reactor power level channels (Linear and Log-N), and the period channel shall be made annually; and TS 4.2.3(4) states, [t]hermal power level shall be verified annually.

However, the purpose of, and difference between, proposed TSs 4.1(7), 4.2.3(3), and 4.2.3(4) is not clear to the NRC staff. Additionally, it is not clear to the NRC staff how proposed TSs 4.1(7), 4.2.3(3), and 4.2.3(4) would be applied for the proposed flux monitoring channels. Explain the purpose of and difference between these surveillance requirements, including specific information regarding any changes to the calibration and verification procedures that apply to the proposed NMP-1000 modules and the proposed Thermo Scientific PPM. Additionally, if necessary, propose revised TSs.

b. SAR Section 7.6.2.1, states that each HMI employs a failsafe watchdog timer that activates trip relays in the scram circuit. Table 3.2.3-1 of proposed TS 3.2.3 (in the proposed TSs submitted March 5, 2019), which stipulates the minimum number of reactor protection system scrams that shall be operable to ensure that the SL is not exceeded, includes LCO TSs for two watchdog timer scrams, item 9, Process Controls Display Watch Dog Timer, and item 10, Drives Controls Display Watch Dog Timer.

Proposed TS 4.2.3(7) would require surveillances of several different scrams, including the watchdog timer scrams. However, the functionality of these timer scrams, and the bases of the LCO and surveillance TSs related to these scrams, is not clear to the NRC staff.

i.

Explain in detail how the watchdog timer activates the trip relays in the scram circuit. Clarify or update information in the SAR, if needed.

ii. Explain how the TS-required surveillance associated with the LCO TSs for the watchdog timer scrams is performed to assure their operability. Include a justification of the annual interval in TS 4.2.3 for verifying operability.

iii. The proposed LCO TSs for the watchdog timer scrams (submitted March 5, 2019), specify that the function of the scrams is a [s]cram for communication loss >10 second. However, the prior version of the proposed license renewal TSs originally submitted with the SAR dated October 20, 2015, specified that these scrams would be required to occur for a communication loss

>1 second. Provide a safety basis justification for changing from 1 to 10 seconds. Additionally, if necessary, propose revised TSs.

RAI-7.15 The regulations in 10 CFR 50.34(b)(2) require that the SAR contain a description and analysis of the SSCs of the facility, with emphasis upon performance requirements; the bases, with technical justification therefor, upon which such requirements have been established; and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as I&C systems shall be discussed so far as they are pertinent.

The guidance in NUREG-1537, Part 2, Section 4, states that ANSI/ANS-15.15-1978 is useful as general guidance for the design, implementation, and evaluation of I&C systems. Clause 5.6 of ANSI/ANS-15.15-1978 states that, [t]he RSS shall include physical features that assure that the proper setpoints are automatically made active or include features that facilitate administrative controls to verify the proper setpoints, or both, when the operating mode of the reactor is changed.

During the Audit, the NRC staff observed that the proposed NMP-1000 Modules have a GA ID and generic password preinstalled for GA use. However, it is not apparent to the NRC staff whether this generic password is removed to limit unauthorized access to the channels once received at UML. Also, it is not evident to the NRC staff that UML has a process to control passwords once the channels are installed for use in the UMLRR (e.g., changing passwords when individuals leave the UMLRR staff or their access privileges change).

Verify that, once the NMP-1000 modules are delivered to UML, the GA ID and password, which is divulged in cleartext within the core module validation report (item No. 12 listed in RAI-7.4.a), will be removed or changed in the equipment prior to its use in the UMLRR. If the GA ID and generic password will not be removed or changed for equipment to be installed for use, explain why not. Additionally, discuss the process UML will use to remove access to NMP-1000 modules when individuals with prior authorized access (i.e., user ID and password) should no longer be authorized access, or explain why this process is not necessary.

Alternatively, discuss why no additional information is required.

RAI-7.16 The guidance in NUREG-1537, Part 1, Section 7.4, states that the information in the SAR should include RPS scram time as established in the accident analysis, and any other requirements to ensure operability.

During the Audit, the NRC staff reviewed the response time constants in Section 1.2, Specifications, in the NMP-1000 Multi-Range Linear Module User Manual, (item No. 14 listed in RAI-7.4.a above). The NRC staff noted that the response time values in the manual appear to be inconsistent with those in the NMP-1000 module specifications provided in SAR Table 7.2 and in the procedure for Test 2.29, EMRT.STP.01, Electrometer - Response Time - Rev A, in the Ethernet Core Module Validation Summary Report (item No. 12 listed in RAI-7.4.a above).

Proposed TS 3.2.1(2) stipulates that the scram time must be less than 1 second from a fully withdrawn control blade position to 80 percent inserted. The 1 second time includes the instrument delay time, as well as the time for the blade to physically drop by gravity to 80 percent inserted. However, given the apparent difference between the response times listed in these documents, it is not clear to the NRC staff how proposed TS 3.2.1(2) would be met, or whether the assumptions of the UMLRR accident analyses (i.e., reactivity transient analyses) are appropriate.

Provide the following, or justify why no additional information is required:

a. Explain why the response times in the user manual (some of which are in units of seconds) appear to be different from those in the SAR and acceptance tests, which are in units of milliseconds. Alternatively, clarify that the response times in the user manual are the same as those in the SAR and acceptance tests, and provide a supporting reference to the user manual.

b. Verify that the response times listed in the SAR and acceptance tests will be used as the acceptance criteria for UMLs acceptance testing of the NMP-1000 channels.

Additionally, verify that the system requirements for the RPS (such as required scram times for both slow and fast scrams) are clearly and correctly identified and are consistent with the assumptions and system requirements in the accident analyses and TSs (SAR Chapters 13 and 14, as supplemented).

RAI-15.1 The Nuclear Energy Innovation and Modernization Act (NEIMA) was signed into law on January 14, 2019. This law, among other things, established a criterion in Section 104(c) of the Atomic Energy Act of 1954, as amended (AEA), for the NRC to use to determine whether a utilization facility is licensed as a commercial or industrial facility (Class 103 license) or a research and development facility (Class 104(c) license).

Section 106 of NEIMA, Encouraging Private Investment in Research and Test Reactors, amends the AEA by adding the following text to the end of Section 104(c):

The Commission is authorized to issue licenses under this section for utilization facilities useful in the conduct of research and development activities of the types specified in section 31 in which the licensee sells research and testing services and energy to others, subject to the condition that the licensee shall recover not more than 75 percent of the annual costs to the licensee of owning and operating the facility through sales of nonenergy services, energy, or both, other than research and development or education and training, of which not more than 50 percent may be through sales of energy.

This criterion is in addition to, and different than, the criterion in 10 CFR 50.22, Class 103 licenses; for commercial and industrial facilities, which states:

A class 103 license will be issued, to an applicant who qualifies, for any one or more of the following: To transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, or use a production or utilization facility for industrial or commercial purposes; Provided, however, That in the case of a production or utilization facility which is useful in the conduct of research and development activities of the types specified in section 31 of the Act, such facility is deemed to be for industrial or commercial purposes if the facility is to be used so that more than 50 percent of the annual cost of owning and operating the facility is devoted to the production of materials, products, or energy for sale or commercial distribution, or to the sale of services, other than research and development or education or training.

(Note: Bold italics used for emphasis.)

The SAR, as supplemented, does not appear to specifically address the criterion in NEIMA regarding the percentage of the annual costs to the licensee of owning and operating the facility that is recovered through sales of nonenergy services, energy, or both, other than research and development or education and training.

Provide a statement identifying whether the percentage of the annual costs of owning and operating the facility recovered through sales of nonenergy services, energy, or both, other than research and development or education and training, is 75 percent or less, and whether the percentage of the annual costs of owning and operating the facility recovered from sales of energy is 50 percent or less.