ML17354A845

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LLC Response to NRC Request for Additional Information No. 277 (Erai No. 8747) on the NuScale Design Certification Application
ML17354A845
Person / Time
Site: NuScale
Issue date: 12/20/2017
From: Rad Z
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-1217-57684
Download: ML17354A845 (23)
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RAIO-1217-57684 December 20, 2017 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.

277 (eRAI No. 8747) on the NuScale Design Certification Application

REFERENCE:

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

277 (eRAI No. 8747)," dated November 03, 2017 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 Questions from NRC eRAI No. 8747:

18-8 18-9 18-10 18-11 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 Steven Mirsky at 240-833-3001 or at smirsky@nuscalepower.com.

Sincerely, Zackary Za Z ckary W. Rad Director, Regulatory Affairs Director NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Demetrius Murray, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 8747 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-1217-57684 :

NuScale Response to NRC Request for Additional Information eRAI No. 8747 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.: 8747 Date of RAI Issue: 11/03/2017 NRC Question No.: 18-8 Regulatory Basis Title 10 of the Code of Federal Regulations (10CFR) Section 52.47(a)(8) requires an applicant for a design certification to provide a final safety analysis report (FSAR) that must include the information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii),

(f)(2)(ix), and (f)(3)(v). Section 10 CFR 50.34(f)(2)(iii) requires an applicant to "Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts.

Chapter 18, Human Factors Engineering, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, and NUREG-0711, "Human Factors Engineering Program Review Model, identify criteria the staff uses to evaluate whether an applicant meets the regulation. The FSAR, Tier 2, Section 18.0, "Human Factors Engineering - Overview," indicates that the HFE program incorporates the applicable guidance provided in NUREG-0711, Revision 3. This regulatory basis applies to all questions in this request for additional information (RAI).

Acceptance Criteria NUREG-0711, Review Criterion 8.4.2(2), states in part, "The applicant should provide an overview of the HSI, covering the technical bases demonstrating that they constitute a state-of-the-art HSI design supporting personnel performance."

Application The Human Factors Engineering Human-System Interface Design Results Summary Report (HD RSR), Section 2.1, "Human- System Interface Design Process Overview," provides an overview of the HSI design process and lists four steps in the HSI design process. Steps 1 and 2 in Section 2.1 are the same.

Evaluation The staff would like to understand if there are indeed four distinct steps, or if there are only three steps.

NuScale Nonproprietary

Question Please revise the HD RSR, Section 2.1, either to delete Step 2 or to identify the activities performed as part of Step 2.

NuScale Response:

There are four distinct steps in Section 2.1 of RP-0316-17619, Human-System Interface Design Results Summary Report. RP-0316-17619 Section 2.1, steps 1 and 2 represent NUREG-0700 Part III Section 11, Workstation Design (Step 2), and Section 12, Workplace Design (Step 1),

and HSI Style Guide sections 3.6 and 3.7, respectively.

RP-0316-17619 section 2.1 step 1 has been revised to "workplace" instead of "work location" to be consistent with NUREG-0700 and the HSI Style Guide.

Impact on DCA:

RP-0316-17619, HFE Human-System Interface Design RSR, has 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 Human-System Interface Design The results of the HSI design analysis, including details of the resulting MCR configuration, are documented in Reference 18.7-2.

The HSI tests and evaluations activities are part of the HSI design analysis, and include HSI inventory and characterization, HSI task support verification, and HSI design verification.

These activities will continue during the detailed design and integration phase to capture the HSI design as it evolves through the verification and validation HFE element.

18.7.4 References RAI 18-6, RAI 18-7, RAI 18-8 18.7-1 NuScale Power, LLC, Human Factors Engineering Human-System Interface Style Guide, ES-0304-1381, Revision 12.

RAI 18-6, RAI 18-7, RAI 18-8 18.7-2 NuScale Power, LLC, Human-System Interface Design Results Summary Report, RP-0316-17619, Revision 01.

Tier 2 18.7-12 Draft Revision 1

Human Factors Engineering Human-System Interface Design Results Summary Report RP-0316-17619-NP Rev. 10 2.0 Implementation 2.1 Human-System Interface Design Process Overview The analyses performed in the early stages of the HFE program are important steps in establishing the inputs to the design requirements for the NuScale HSIs. The HSI design inputs that are analyzed and/or developed include the following:

  • operating experience review (OER)
  • functional requirements analysis and function allocation (FRA/FA)
  • task analysis (TA)
  • staffing and qualifications (S&Q)
  • treatment of important human actions (IHAs)
  • concept of operations
  • l&C systems design
  • alarm management
  • system requirements
  • HSI Style Guide Once the inputs are established, the design effort follows the NuScale HSI process steps listed below when designing the MCR, conceptual workstations, and screen-based HSIs needed to complete the design effort.
1. Follow the appropriate chapters of the NuScale HSI Style Guide needed to establish a safe, user-friendly workplace location.
2. Follow the appropriate chapters of the NuScale HSI Style Guide needed to establish safe, user-friendly workstations.
3. Design and develop the HSI needed to accomplish safe and reliable operation of the plant.
4. Test and evaluate the HFE/HSI design of the simulator and products developed to support SPV testing The HSI design products are the physical HSI screens, the embedded procedure functionality, and the plant notification functionality maintained within the simulator control room hardware and software. Examples and illustrations of these results are provided in Section 4.0 of this report.

© Copyright 20176 by NuScale Power, LLC 7

Response to Request for Additional Information Docket No.52-048 eRAI No.: 8747 Date of RAI Issue: 11/03/2017 NRC Question No.: 18-9 Acceptance Criteria NUREG-0711, Section 13.3, "Applicant Products and Submittals," states, "The applicant should provide for review an IP for monitoring human performance after the plant becomes operational."

Application The Human Factors Engineering Program Management Plan (HFE PMP), Table 4-2, "Human factors engineering element documentation", states, "Human performance monitoring is the responsibility of a COL applicant. No implementation plan or RSR [results summary report] is submitted as part of design certification application." FSAR Tier 2, Section 18.12, "Human Performance Monitoring," includes COL Item 18.12-1 for the COL applicant to develop the human performance monitoring program.

However, the HFE PMP, Section 6.11, "Human Performance Monitoring," says, " An HPM IP is produced as part of the NuScale HFE program as guidance for the COL applicant over the life of the plant."

Evaluation The staff would like to understand whether NuScale intends to submit a human performance monitoring implementation plan as part of the design certification application.

Question Please clarify whether NuScale intends to submit a human performance monitoring implementation plan as part of the design certification application.

If NuScale intends to provide a human performance monitoring implementation plan, then please state when it will be provided and remove COL Item 18.12-1 from the DCD.

If NuScale does not intend to provide a human performance monitoring implementation plan, then please revise the HFE PMP, Section 6.11, to align with the information in the FE PMP, NuScale Nonproprietary

Table 4-2.

NuScale Response:

NuScale does not intend to submit a human performance monitoring implementation plan.

RP-0914-8534, HFE Program Management Plan, Section 6.11 has been revised to be consistent with Table 4-2 "Human factors engineering element documentation."

Changes to the HFE Program Management Plan are attached.

Impact on DCA:

RP-0914-8534, Human Factors Engineering Program Management Plan, has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Human Factors Engineering Program Management Plan RP-0914-8534-NP Rev. 34 6.10 Design Implementation The NuScale HFE program includes design implementation. The DI IP describes the DI process and includes the output documentation requirements. No RSR is required for this element since conformance of the as-built design to the verified and validated design is confirmed by an ITAAC.

DI demonstrates that the design that is implemented accurately reflects the design that has been verified and validated during human factors V&V. If the DI program element identifies differences such as site-specific aspects that were not included in V&V or design changes that occur after V&V, those differences are evaluated to determine any impact to the analysis results from all previous HFE program elements, including V&V.

While successful ISV marks the end of human factors V&V, the HSI design continues to be challenged during the operator training program. Any Priority 3 HEDs generated during V&V that are determined to require resolution and any HEDs generated after completion of V&V are resolved during DI.

6.11 Human Performance Monitoring Human performance monitoring (HPM) begins after DI is completed and continues for the life of the plant. HPM is intended to detect degradation in operator performance compared to the performance observed during integrated system validation. Degradation may be due to many factors that occur over the life of the plant, including changes in personnel, changes in plant culture, changes in training methods, or changes in the HSI design itself. The HPM program includes

  • monitoring and investigating perceived or documented reduced human performance.
  • analyzing causes of reduced human performance.
  • developing corrective action plans for improvement.
  • maintaining a culture of continuous monitoring of human performance through operating experience review.
  • training and qualification.
  • change management (modification process, configuration management).
  • use of the plant simulator.
  • independent reviews and audits.

HPM is a catalyst for corrective actions over the life of the plant; the COL applicant manages its own corrective actions program.

HPM is a responsibility of the COL applicant. An HPM IP is produced as part of the NuScale HFE program as guidance for the COL applicant over the life of the plant.No implementation plan or RSR is submitted as part of the design certification application.

© Copyright 20162017 by NuScale Power, LLC 34

Response to Request for Additional Information Docket No.52-048 eRAI No.: 8747 Date of RAI Issue: 11/03/2017 NRC Question No.: 18-10 Acceptance Criteria NUREG-0711, Review Criteria 6.4(4), says, "The applicants staffing analysis should determine the number and qualifications of operations personnel for the full range of plant conditions and tasks, including operational tasks (under normal, abnormal, and emergency conditions), plant maintenance, plant surveillance, and testing."

Relevant Background Information On June 23, 2016, and August 30, 2016, Category 1 public meetings were held at the NRC Headquarters with representatives of the NRC staff and NuScale to discuss the NRC staffs response to the letter from NuScale to the NRC dated June 8, 2016, titled, Regulatory Process for Addressing Licensed Operator Staffing Regulations in the NuScale Design Certification, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16168A463). The letter states, "NuScale anticipates licensees referencing the NuScale design to use control room configurationsnumbers of operators per reactor and number of reactors operated from a control roomthat differ from existing large reactors and the requirements stated at 10 CFR 50.54(m). While the requirements for minimum licensed operator staffing are a license condition, and thus applicable to a licensee, NuScale will seek to resolve the matter in the NuScale design certification (DC)."

The NRC staff response was documented in Enclosure 1 of the meeting summary dated September 13, 2016 (ADAMS Accession No. ML16252A258). In Enclosure 1, the staff listed text to be included in the design certification application, including "a statement that the minimum staffing requirements are located in the Design Certification Rule Part 52 Appendix" in DCD Tier 1, section for human factors engineering, and a discussion of why an exemption is not necessary for NuScale in Part 7 of the design certification application. Also, Enclosure 1 stated NuScale should include a proposed staffing table and appropriate table notes in Part 7.

Application

1. The staff reviewed DCD Tier 1, Section 3.15 , "Human Factors Engineering," and did not find a statement that the minimum staffing requirements are located in the Design Certification Rule Part 52 Appendix.

NuScale Nonproprietary

2. The staff reviewed Part 7, Section 6, "The staff reviewed Part 7, Section 6, "10 CFR 50.54(m), Control Room Staffing," and did not find a discussion of why an exemption is not necessary for NuScale.
3. The staff reviewed the staffing table and table notes provided in Part 7, Section 6, "The staff reviewed Part 7, Section 6, "10 CFR 50.54(m), Control Room Staffing." Note 2 on the proposed staffing table includes a statement that "a nuclear power unit is considered to be operating when it is fueled, in an operating bay, and has the ability to communicate with a support system as defined by the unit's technical specifications."

Part 4, "Generic Technical Specifications," Section 5.2.2, "Facility Staff," of the design certification application includes a similar table with one table note that says, "For the purposes of this table, a MODULE is considered to be operating when it is in MODE 1, 2, or 3."

4. The proposed staffing table provided in Part 7, Section 6.1.3, "Requested Action," says that the minimum number of operators is the same regardless of whether there are one to twelve units operating. Part 6, Section 6.2.1, "Technical Basis," says, "The NuScale-proposed staffing requirements are consistent with NUREG-0800, Chapter 18; NUREG-0711, Revision 3; NUREG-1791; and NUREG/CR-6838, February 2004 (endorsed in NUREG-0711) for the review criteria of plant staffing levels that require an exemption from 10 CFR 50.54."

RP-0516-49116, "Control Room Staffing Plan Validation Results," describes the method used to perform the staffing plan validation (SPV) described in NUREG-1791 and the results of results of NuScale' SPV. RP-0516-49116, Abstract , says, This report has been developed to describe the results of staffing plan validation testing performed to evaluate licensed operator workload in challenging high workload situations within a NuScale 12-unit control room environmentThe results of the analysisconfirm that up to 12 NuScale power modules and the associated plant facilities may be operated safely and reliably by a minimum staffing contingent of three licensed reactor operators and three licensed senior reactor operators from a single control room during normal, abnormal, and emergency conditions.

5. Part 4, "Generic Technical Specifications," Section 5.2.2.c. states, a licensed reactor operator or senior reactor operator shall be present at the controls at all times. Part 7, Section 6.1.3(3) also states, a licensed reactor operator or senior reactor operator shall be present at the controls at all times.
6. Part 4, "Generic Technical Specifications," Section 5.2.2 uses the word MODULES in the proposed minimum staffing table. The proposed staffing table in Part 7, Section 6 uses the phrase nuclear power units.

Evaluation

1. The staff would like to understand why the DCD Tier 1, Section 3.15 does not include a statement that the minimum staffing requirements are located in the Design Certification Rule Part 52 Appendix.
2. The staff would like to understand why Part 7, Section 6 does not include a discussion NuScale Nonproprietary

of why an exemption is not necessary for NuScale.

3. The staff would like to understand the phrase in Note 2 of the proposed staffing table in Part 7, Section 6 that says, "...has the ability to communicate with a support system as defined by the unit's technical specifications," and how this statement relates to Note 1 of the staffing table included in Section 5.2.2 of Part 4, "Generic Technical Specifications, which says, For the purposes of this table, a MODULE is considered to be operating when it is in MODE 1, 2, or 3."
4. For the SPV, 12 units were simulated in the control room simulator and high workload scenarios were administered in order to demonstrate that the proposed minimum staffing level was acceptable under high workload conditions. However, human performance can also be affected negatively when workload is too low. The staff would like to understand whether workload analyses have been or will be done to evaluate whether workload is above an acceptable lower limit when fewer than 12 units exist at a plant.
5. The staff would like to understand what is considered at the controls for a NuScale plant. As defined in 10 CFR 50.2, Definitions, the word controls when used with respect to nuclear reactors means apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.
6. The staff would like to understand whether units and MODULES are synonymous.

Questions

1. Please explain why DCD Tier 1, Section 3.15 does not include the statement, or revise DCD Tier 1, Section 3.15, to include a statement that the minimum staffing requirements are located in the Design Certification Rule Part 52 Appendix.
2. Please explain why Part 7, Section 6 does not include a discussion of why an exemption is not necessary for NuScale, or revise Part 7, Section 6 to include a discussion of why an exemption is not necessary for NuScale.
3. Please explain what is meant by "support systems as defined by the unit's technical specifications," and how this statement relates to the note provided with the staffing table included in Section 5.2.2 of Part 4, "Generic Technical Specifications."
4. Please explain whether low workload has been or will be evaluated for the case where fewer than 12 units are operated from a single control room in order to justify that the minimum staffing level is appropriate for one to 12 modules.
5. Please explain what is considered at the controls for a NuScale plant.
6. Please explain whether units and modules are synonymous.

NuScale Nonproprietary

NuScale Response:

Response to sub-question 1 The alternative minimum licensed operator staffing requirements requested by NuScale will be a condition of licenses referencing the NuScale design certification. Furthermore, a licensee referencing the NuScale design certification will be required to "incorporate by reference, as part of its application" the appendix certifying the NuScale design, which will include the minimum staffing requirements (see, e.g., 10 CFR 52, App. E,Section IV.A.1). The proposed staffing requirements are also reflected in the NuScale Standard Technical Specifications, TS 5.2.2, which is appropriate for an operational requirement. These provisions assure that a licensee is aware of the minimum licensed operator staffing requirements and provide appropriate means of change control. Reference to operator staffing requirements in Tier 1 would not further contribute to design standardization and may cause confusion because such statement is atypical of Tier 1 information.

Response to sub-question 2 The regulations addressed by Part 7 Section 6 are conditions of operating licenses issued under Part 50 or Part 52. Therefore, NuScale is unable to request an exemption from 10 CFR 50.54 because NuScale is not a licensee and has not applied for an operating license. Part 7 Section 6 has been updated to explicitly state this. Changes are attached and will be incorporated in the DCA Revision 1 submittal.

Response to sub-question 3 The phrase "support systems as defined by the unit's technical specifications" was included in Part 7 in error. The inconsistency between note 2 of the proposed staffing table in Part 7, Section 6 and note 1 of the staffing table in Section 5.2.2 of Part 4 has been resolved, as shown in the attached markups.

Response to sub-question 4 NuScale does not plan to specifically evaluate low work load in the case where fewer than 12 units are in operation. The NuScale design has few simple, passive safety systems, few support systems, and does not rely on AC or DC power for nuclear safety. The NuScale design does not rely on operator actions to mitigate design basis events (DBEs), and only seven operator actions are identified in the full-power internal events probabilistic risk assessment for beyond-design-basis events (BDBEs), six of which are performed in the Main Control Room. PRA fault tree analysis was conducted for sequences that lead to core damage. All sequences leading to core damage have very low probabilities, and all are BDBEs. Operator errors of both omission and commission were analyzed in FSAR Chapters 15 and 19. The analysis concluded that failure to take action cannot make the consequences more severe than the bounding Chapter 15 analysis. (Reference FSAR Chapter 15, Section 15.0.0.5 and Section 15.0.0.6.4). When NuScale Nonproprietary

analyzing Chapter 19 BDBEs, considering that all human actions fail, the NuScale core damage frequency (CDF) would rise to approximately 3.0E-8 per module critical year. For these reasons, periods of low workload and the resultant potential lapse of situational awareness do not impact the plant's designed level of safety.

Response to sub-question 5 10 CFR 50.2, Definitions, defines the word controls when used with respect to nuclear reactors to mean "apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor." The technical specifications section 5.2.2 states that: "A senior reactor operator license shall be in the control room at all times. In addition to this senior reactor operator, a licensed reactor operator or senior reactor operator shall be present at the controls at all times." Manipulations that can directly affect reactivity or power level of the reactor can be performed from either the stand-up Module Control System (MCS) displays or at any of the Plant Control System workstations in the main control room. Therefore an operator is "at the controls" when working at the Shift Tech Advisor workstation, at the Control Room Supervisor workstation, at the Shift Manager workstation, or anywhere between the Control Room Supervisor workstation and the semi-circle of MCS displays. Manipulations that can directly affect reactivity or power level of the reactor cannot be performed from the Safety Display and Indication (SDI) display panels.

Response to Sub-question 6 There is a distinction between units and modules. "MODULE" refers to the structures, systems, and components that form a single integrated reactor, reactor coolant, and containment designed to be disconnected from the power generation and support systems. "UNIT" refers to the MODULE and all support systems that are not shared with other units (e.g. main steam, main feed, and main turbine). To clarify the discrepancy between Part 7, Section 6 and Part 4, Generic Technical Specifications, Section 5.2.2, the definition of "MODULE" was deleted in Generic Technical Specifications Section 1.1, Definitions, and the word "module(s)" was replaced by "unit(s)" in Generic Technical Specifications Section 5.2.2, Facility Staff. Changes to Part 4, Generic Technical Specifications, are attached and will be incorporated in the DCA Revision 1 submittal.

Impact on DCA:

Part 4, Section 5.2.2 and Part 7, Section 6 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Organization 5.2 5.2 Organization 5.2.2 Facility Staff (continued)

a. The minimum licensed operator staffing shall be:

Number of Reactor Senior Reactor unitsMODULES Operator Operator Operating(1)

None 2 1 One to twelve 3 3 (1) For the purpose of this table, a unitMODULE is considered to be operating when it is in MODE 1, 2, or 3.

b. A person holding a senior reactor operator license for all fueled units at the site who is assigned responsibility for overall plant operation shall be onsite at all times when there is fuel in any unitMODULE.
c. A senior reactor operator license shall be in the control room at all times.

In addition to this senior reactor operator, a licensed reactor operator or senior reactor operator shall be present at the controls at all times.

d. Shift crew composition may be less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
e. A radiation protection technician shall be on site when fuel is in any unitMODULE. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
f. The operations manager or assistant operations manager shall hold an SRO license.
g. An individual shall provide advisory technical support to the facility operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the facility. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

NuScale 5.2-2 Draft Revision 1.0

Exemptions 10 CFR 50.54(m), Control Room Staffing

6. 10 CFR 50.54(m), Control Room Staffing 6.1 Introduction and Request 6.1.1 Summary RAI 18-10 10 CFR 50.54(m) specifies minimum licensed operator staffing requirements and responsibilities as a license condition on operating licenses. These requirements do not address a design with more than three units on a site or more than two units operated from a single control room. Further, licensee decisions regarding licensed operator staffing, including the number, composition, and qualifications of licensed personnel are more appropriately based on features unique to the design rather than the existing large, light water reactor-based staffing levels. Because 10 CFR 50.54(m) is applicable only to a licensee, an exemption from the regulations is not appropriate for a design certification applicant. Therefore, NuScale Power, LLC is requesting that design-specific staffing requirements be incorporated in the design certification to be applicable to licensees referencing the NuScale Power Plant design certification, in lieu of the current requirements of 10 CFR 50.54(m).

6.1.2 Regulatory Requirements The introduction to 10 CFR 50.54 states:

The following paragraphs of this section, with the exception of paragraphs (r) and (gg),

and the applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor operating license issued under this part. The following paragraphs with the exception of paragraph (r), (s), and (u) of this section are conditions in every combined license issued under part 52 of this chapter, provided, however, that paragraphs (i) introductory text, (i)(1), (j), (k), (l), (m), (n), (w), (x), (y), (z), and (hh) of this section are only applicable after the Commission makes the finding under § 52.103(g) of this chapter.

10 CFR 50.54(m) states:

(m) (1) A senior operator licensed pursuant to part 55 of this chapter shall be present at the facility or readily available on call at all times during its operation, and shall be present at the facility during initial start-up and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling, or as otherwise prescribed in the facility license.

(2) Notwithstanding any other provisions of this section, by January 1, 1984, licensees of nuclear power units shall meet the following requirements:

(i) Each licensee shall meet the minimum licensed operator staffing requirements in the following table:

Part 7 6-1 Draft Revision 1

Exemptions 10 CFR 50.54(m), Control Room Staffing V. Applicable Regulations C. A licensee referencing this appendix is exempt from portions of the following regulations:

1. Paragraph (m) of 10 CFR 50.54Conditions of licensescodified as of

[date of NuScale Power Plant design certification]. In place, the following requirements shall be conditions of such licenses:

a. A senior operator licensed pursuant to part 55 of this chapter shall be present at the facility or readily available on call at all times during its operation, and shall be present at the facility during initial start-up and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling, or as otherwise prescribed in the facility license.
b. Licensees shall meet the following requirements:

(1) Each licensee shall meet the minimum licensed operator staffing requirements in the following table:

RAI 18-10 Minimum Requirements1 Per Shift for On-Site Staffing of NuScale Power Plants by Operators and Senior Operators Licensed Under 10 CFR Part 55 Number of nuclear power units Position One to twelve units operating2 One control room None Senior Operator 1 Operator 2 One to twelve Senior Operator 3 Operator 3 1

Temporary deviations from the numbers required by this table shall be in accordance with criteria established in the unit's technical specifications.

2 For the purpose of this table, a nuclear power unit is considered to be operating when it is fueled, in an operating bay, and has the ability to communicate with a support system in MODE 1, 2, or 3 as defined by the unit's technical specifications.

(2) Each licensee shall have at its site a person holding a senior operator license for all fueled units at the site who is assigned responsibility for overall plant operation at all times there is fuel in any unit.

(3) When a nuclear power unit is fueled, in an operating bay, and has the ability to communicate with a support system, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator a licensed operator or senior operator shall be present at the controls at all times. In addition to the senior operator and licensed operator or senior operator present at the controls, a licensed operator or Part 7 6-3 Draft Revision 1

Response to Request for Additional Information Docket No.52-048 eRAI No.: 8747 Date of RAI Issue: 11/03/2017 NRC Question No.: 18-11 Acceptance Criteria NUREG-0711, Section 11.3, Applicant Products and Submittals, says, The applicant should provide either an IP [implementation plan] or a completed RSR [results summary report]. If the applicant submits an IP, it should describe the complete methodology for conducting V&V At a minimum, the RSR should include the following: a description of the methodology, if an NRC approved IP was not usedIf the methodology was described in an IP that the NRC staff previously reviewed, the contents of the RSR should be consistent with the approved methodology and the applicant should discuss the rationale for any deviations from it.

Relevant Background Information The public meeting summary dated April 11, 2016 (ADAMS Accession No. ML16060A220),

summarizes the discussion between the NRC staff and NuScale on human factors engineering topics. In this meeting the staff said they would docket and commence the review of the NuScale design certification application if NuScale submitted an IP for the verification and validation (V&V) activity, and the V&V RSR should be submitted no later than the start of Phase 4 of the DC review.

Application NuScale provided a V&V IP with the design certification application. FSAR Tier 2, Chapter 1, Table 1.6-2, identifies the human factors engineering RSRs and IPs that are incorporated by reference in the FSAR Tier 2, Chapter 18. Table 1.6-2 identifies that the V&V IP is incorporated by reference in the FSAR Tier 2, Chapter 18.

However, FSAR Tier 2, Section 18.1.3.5, Human Factors Engineering Documentation, says, An IP is prepared for each HFE elementand submitted for NRC review. The IP for a given element describes the methodology for conducting that element. Upon completion of the associated HFE activities, RSRs are prepared for the following HFE elements:

human factors verification and validation (Section 18.10). The RSRs for these elements contain the results and the latest methodology, and supersede the previously-submitted IPs.

NuScale Nonproprietary

Evaluation The staff is currently reviewing the V&V methodology in the V&V IP. The staff would like to understand whether NuScale intends to supersede the V&V IP when the verification and validation (V&V) RSR is submitted.

If NuScale does intend to supersede the V&V IP when the V&V RSR is submitted, then the V&V IP will not be part of the FSAR. Therefore, FSAR Tier 2, Chapter 1, Table 1.6-2, will need to be revised to remove the V&V IP as a document that is incorporated by reference in the FSAR.

Also, the V&V RSR will need to include a description of the V&V methodology at a level of detail that will allow the staff to determine how the review criteria in NUREG-0711, Section 11, have been addressed.

If NuScale does not intend to supersede the V&V IP when the V&V RSR is submitted, then the V&V IP will remain part of the FSAR. The V&V RSR will not need to include a detailed description of the V&V methods unless the methods actually used during V&V activities differ from those described in the V&V IP.

Additionally, FSAR Tier 2, Chapter 1, Table 1.6-2, will need to be revised to include the V&V RSR.

Question Please state whether NuScale intends to supersede the V&V IP when the V&V RSR is submitted.

NuScale Response:

NuScale will not supersede the Human Factors Verification and Validation (V&V) implementation plan (IP) when the V&V results summary report (RSR) is submitted. NuScale submitted the V&V IP in order that NRC staff better understand the methodology used by NuScale during performance of V&V activities described in NUREG-0711. NuScale will not include a detailed description of the V&V methodology in the V&V RSR unless the methods actually used during V&V activities differ from those described in the V&V IP.

RP-0914-8534, Human Factors Engineering (HFE) Program Management Plan (PMP), Section 4.5 has been revised to state: "Since the human factors verification and validation RSR will be completed after the initial DCA submittal, the human factors verification and validation IP will remain a standalone document. As a result, the human factors verification and validation RSR will not contain a methodology section but will simply reference the IP."

FSAR Table 18.1-1, Human Factors Engineering Program and Design Activity Milestones, NuScale Nonproprietary

indicates that both the V&V IP and the V&V RSR will be submitted as part of the design certification application (DCA). The V&V RSR is modified by a Note 1 as follows: "In accordance with Reference 18.1-2, verification and validation RSR is to be submitted prior to start of Phase 4 of the NRCs review of the design certification application."

FSAR Section 18.1.3.5, Human Factors Engineering Documentation, lists all of the RSRs that are to be submitted as part of the DCA. Subsequently, there is a sentence "The RSRs for these elements contain the results and the latest methodology, and supersede the previously-submitted IPs." This statement is accurate for all of the RSRs with the exception of the V&V RSR. FSAR Section 18.1.3.5 has been revised to clarify that the V&V IP is not superseded by the V&V RSR, consistent with Section 4.5 of RP-0914-8534, HFE PMP.

A conforming change was made to FSAR Section 18.2.4.

Impact on DCA:

FSAR Sections 18.1.3.5 and 18.2.4 and RP-0914-8534, HFE Program Management Plan, 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 Human Factors Engineering Program Management

  • staffing and qualifications (Section 18.5)
  • treatment of important human actions (Section 18.6)
  • HSI design (Section 18.7)
  • human factors verification and validation (Section 18.10)

RAI 18-11 The RSRs for these elements, with the exception of the human factors verification and validation RSR, contain the results and the latest methodology, and supersede the previously-submitted IPs. Since the human factors verification and validation RSR will be completed after the initial DCA submittal, the human factors verification and validation IP will remain a standalone document. As a result, the human factors verification and validation RSR will not contain a methodology section but will simply reference the IP. The RSRs contain sufficient detail to demonstrate that the results were derived from implementing the methodology. The RSR scope is in accordance with the applicable guidance of NUREG-0711.

The HFE documents that support the design are quality records and are retained in accordance with the quality assurance program (Section 17.5). The HFE documentation includes design verification checklists, HFEITS records (see Section 18.1.4), HFE element IPs, RSRs, and applicable documentation identified in the IPs and RSRs.

18.1.3.6 Subcontractor Human Factors Engineering Efforts Subcontractors may be utilized in the HFE program. The HFE team verifies that any subcontractor performing HFE activities is properly trained and complies with the quality assurance program and the applicable subordinate plans and procedures. The quality assurance organization verifies that the subcontractors conduct work in accordance with the quality assurance program or the subcontractor's quality assurance programs, as approved and contracted.

18.1.4 Tracking Human Factors Engineering Issues 18.1.4.1 Availability of Human Factors Engineering Issue Tracking System If identified HFE issues cannot be immediately resolved, they are included and tracked in the HFEITS database, which is available to the HFE team members. HFE issues may include recognized industry HFE issues, HEDs identified during HFE design, and issues identified throughout the life cycle of the HFE program. Details on the HFEITS process are contained in Reference 18.1-1.

18.1.4.2 Human Factors Engineering Issue Tracking Method Identified HFE issues that cannot be immediately resolved are entered into the HFEITS database and assigned a unique tracking number. Supporting documentation in electronic format is attached to the database item. The issue is screened and evaluated for potential degradation in human performance. Issues that are found to not degrade human performance are either closed or transferred to more appropriate corrective action processes.

Tier 2 18.1-7 Draft Revision 1

NuScale Final Safety Analysis Report Operating Experience Review

  • important human actions The fundamental NuScale design features eliminate the potential for the operating experiences encountered in the commercial nuclear power industry such as
  • natural circulation within the primary system eliminating the potential for operating experience related to reactor coolant pumps, motors, and seals.
  • the integrated reactor module design eliminating piping, welds, and valves associated with an external pressurizer and steam generators.
  • the small containment and evacuated annulus eliminating the need for thermal insulation around the reactor vessel, reducing GSI-191 concerns.
  • simple, passive safety systems that transition to a state that meets their safety function on a loss of power, eliminating the need for safety-related AC or DC power and their associated backup and support systems.

The issues identified from the reviews as being applicable to NuScale Power Plant design were incorporated into the design. The following are examples of significant enhancements to NuScale Power Plant design resulting from the OER (with the associated type of OER noted parenthetically):

  • minimizing and prioritizing alarms in the control room, thereby reducing the likelihood of alarm avalanche (related HSI technology)
  • providing improved methods for allocation of information across workstations (issues identified by plant personnel)
  • providing diverse HSI capabilities to allow operators to cope with postulated failures or degradation of the normally used HSIs (issues identified by plant personnel)

The results of reviews of each of the five OER areas are documented in Reference 18.2-1.

18.2.4 References 18.2-1 NuScale Power, LLC, Human Factors Engineering Operating Experience Review Results Summary Report, RP-0316-17614, Revision 0.

RAI 18-11 18.2-2 NuScale Power, LLC, Human Factors Engineering Program Management Plan, RP-0914-8534, Revision 43.

Tier 2 18.2-7 Draft Revision 1

Human Factors Engineering Program Management Plan RP-0914-8534-NP Rev. 34 HFE and Design Activities Activity Milestones RSR)

SSC Design X X SSC Testing X 4.5 Human Factors Engineering Documentation HFE documents that support design are quality records and retained in accordance with the NuScale QMP (Reference 8.2.5). All such documentation is available for review upon request. HFE documentation includes design verification checklists, HFEITS records, documentation identified in the HFE element technical reports (e.g., RSRs, guides, and training programs), and information stored in the HFE database.

Table 4-2 provides a tabulation of the IPs and the RSRs submitted. The resultsmethodologies of the HFE activities described in each elements implementation plan are summarized in the associated RSR, with the exception of the V&V RSR. In addition to results of the analyses, the RSRs (References 8.2.8 through 8.2.13) contain the latest methodology used in the analysis, and supersede the methodology described in the associated IPs that had been previously submitted. Since the human factors verification and validation RSR will be completed after the initial DCA submittal, the human factors verification and validation IP will remain a standalone document. As a result, the human factors verification and validation RSR will not contain a methodology section but will simply reference the IP.

Unless otherwise noted, the contents of the RSRs are in accordance with the applicable guidance of NUREG-0711.

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