ML15120A254

Enclosure 1: Response to Request for Additional Information

ML15120A254
Person / Time
Site:	SHINE Medical Technologies
Issue date:	04/10/2015
From:	SHINE Medical Technologies
To:	Office of Nuclear Reactor Regulation
Shared Package
ML15120A248	List: ML15120A241 ML15120A252 ML15120A253 ML15120A254
References
SMT-2015-012
Download: ML15120A254 (28)

Text

SHINE MEDICAL TECHNOLOGIES, INC.

SHINE MEDICAL TECHNOLOGIES, INC. APPLICATION FOR CONSTRUCTION PERMIT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION The NRC staff determined that additional information was required (Reference 1) to enable the continued review of the SHINE Medical Technologies, Inc. (SHINE) application for a construction permit to construct a medical isotope facility (References 2 and 3). The following information is provided by SHINE in response to the NRC staffs request.

GENERAL INFORMATION REQUEST RAI G-3 Title 10 of the Code of Federal Regulations (10 CFR) Paragraph 50.33(e), requires that an application shall state, in part, the class of license applied for.

On October 17, 2014, the NRC published a direct final rule in the Federal Register (FR), which went into effect on December 31, 2014 (79 FR 62329). This rulemaking modified the definition of utilization facility in 10 CFR 50.2, Definitions, to include an accelerator-driven subcritical operating assembly used for the irradiation of materials containing special nuclear material and described in the application assigned docket number 50-608.

In Section 2.4, Class of License, of Enclosure 4 to SHINE letter dated March 26, 2013 (ADAMS Accession No. ML13088A192), SHINE stated, in part, that this application is for the

[construction permit] for a production facility under 10 CFR 50. Additional future applications will be for the production facility operating license under 10 CFR 50 In light of the recent rulemaking modifying the definition of utilization facility in 10 CFR 50.2 to include SHINEs irradiation units, additional information is needed for the NRC staff to determine the adequacy of the description of the class of license applied for in the SHINE construction permit application, as required by 10 CFR 50.33(e).

a. Provide additional information clarifying the types and numbers of facilities (i.e., production and/or utilization facilities) that the SHINE construction permit application supports.

b. Provide information describing any necessary modifications to the SHINE construction permit application to meet the regulatory requirements for any type of facility not previously considered, as applicable.

c. Relate the use of SHINE-specific facility designations (e.g., Production Facility, Medical Isotope Production Facility, Radioisope Production Facility, and Irradiation Facility) to the 10 CFR 50.2 definitions of production facility and utilization facility.

Page 1 of 28

SHINE Response

a. The SHINE Construction Permit Application supports one production facility and eight utilization facilities.

b. As a result of rulemaking modifying the definition of utilization facility in 10 CFR 50.2 to include SHINEs irradiation units, the SHINE description of the type of license being applied for requires modification. The following description of the type of license being applied for supersedes the description previously provided in Section 2.4 of Enclosures 3 and 4 of Reference (2):

2.4 Class of License The class of license SHINE is applying for will be a Class 103 license per 10 CFR 50.22 (for commercial and industrial facilities). This application is for the construction permit for one production facility and eight utilization facilities under 10 CFR 50. Additional future applications will be for the operating license for one production facility and eight utilization facilities under 10 CFR 50, receipt, possession and use of source material under 10 CFR 40, byproduct material under 10 CFR 30, and special nuclear material under 10 CFR 70. SHINE expects to request an operating license (OL) for a term of 30 years.

In addition, SHINE has revised Section 15.2 of the PSAR to state that future applications will be for the operating license for one production facility and eight utilization facilities under 10 CFR 50. A mark-up of the PSAR changes is provided in Attachment 1. The non-public (proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 2. The public (non-proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 3.

c. SHINE uses the terms production facility, production facility building, and medical isotope production facility throughout the PSAR to describe the SHINE facility, as shown on Figure 1.3-2 of the PSAR.

The term production facility, as defined in 10 CFR 50.2, relates to the radioisotope production facility (RPF) portion of the SHINE facility. The term utilization facility, as defined in 10 CFR 50.2, relates to each of the eight irradiation units (IUs) located within the irradiation facility (IF) portion of the SHINE facility.

Page 2 of 28

CHAPTER 3 - DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS Section 3.4 - Seismic Damage RAI 3.4-8 NUREG-1537, Part 1, Section 3.4, Seismic Damage," states that the applicant should include information on the facility seismic design to provide reasonable assurance that the reactor could be shut down and maintained in a safe condition or that the consequences of accidents would be within the acceptable limits in the event of potential seismic events.

NUREG-1537, Part 2, Section 3.4, Seismic Damage, states that the reviewer should find sufficient information to conclude that the design to protect against seismic damage provides reasonable assurance that the facility structures, systems, and components will perform the necessary safety functions described and analyzed.

While SHINEs response to RAI 3.4-1(b) provided additional information on free field site response analyses using the SHAKE2000 computer program, the response did not include a reference for the version of the SHAKE2000 computer program used in SHINEs seismic analysis.

Provide the reference for the version of the SHAKE2000 computer program used in SHINEs seismic analysis.

SHINE Response Version 3.5 of the SHAKE2000 computer program was used in the SHINE seismic analysis.

RAI 3.4-9 NUREG-1537, Part 1, Section 3.4, Seismic Damage, states that in order to verify that seismic design functions are met, the applicant should give the technical specifications necessary to ensure operability, testing, and inspection of associated systems, including instrumentation and controls.

NUREG-1537, Part 2, Section 3.4, Seismic Damage, states that the reviewer find that the surveillance activities proposed provide reasonable assurance that the safety-related functions of the structures, systems, and components (SSCs) that are required to respond to, or mitigate the consequences of, seismic damage to the facility will be maintained.

SHINE PSAR, Section 3.4.4, Seismic Instrumentation, states that the seismic instrumentation operates during SHINE facility operation. The maintenance and repair procedures will keep the maximum number of instruments in service. The in-service testing provisions include periodic channel checks, and the capability for in-place functional testing.

However, in response to RAI 3.4-6, SHINE states that seismic instrumentation will not be installed at the SHINE facility, citing no regulatory requirement for such instrumentation under 10 CFR Part 50 or Part 100. SHINE instead states that procedures will be used to systematically assess the operability and functionality of the plant SSCs. The magnitude of the event will be determined using information available from the U.S. Geological Survey (USGS) or other authoritative source.

Page 3 of 28

While the USGS and other authoritative sources could provide information such as the epicenter location or focal mechanism for a felt earthquake at the SHINE facility, they cannot provide the acceleration time histories or response spectra experienced at the facility needed to conduct post-earthquake quantified evaluations and re-qualifications of the facility Seismic Category I and Seismic Category II SSCs. Additional information is needed for NRC staff to determine that the safety-related functions of the structures, systems, and components that are required to respond to, or mitigate the consequences of, seismic damage to the facility will be maintained.

Provide a discussion of the methodology that will be used to develop procedures for post-earthquake evaluations and re-qualifications of Seismic Category I and Seismic Category II SSCs without onsite seismic instrumentation recording information such as acceleration time histories and response spectra.

SHINE Response Seismic instrumentation is not required as referenced under Section IV(a)(4) of Appendix S to 10 CFR 50 or Section VI(a)(3) of Appendix A to 10 CFR 100, since the SHINE facility is not a nuclear power plant. However, SHINE has decided that for asset protection and business operations purposes, nonsafety-related seismic instrumentation could assist with the timely determination of the accelerations experienced at the plant following a seismic event.

Therefore, SHINE will acquire and install a nonsafety-related seismic monitoring system to help establish the acceptability of continued operation of the plant following a seismic event.

This system will provide acceleration time histories or response spectra experienced at the facility to assist in verifying that structures, systems, and components (SSCs) important to safety at the SHINE facility can continue to perform their safety functions. SHINE will provide an explanation of the data retrieval and processing procedure(s) and state whether or not a separate computer is required to view the digitized acceleration time histories and generate response spectra in the Final Safety Analysis Report (FSAR). An Issues Management Report (IMR) has been initiated to track the inclusion of a description of the seismic monitoring system, as discussed in this response, in the FSAR.

Section 3.5 - Systems and Components RAI 3.5-6 SHINE PSAR, Section 3.5.1, Classification of Systems and Components Important to Safety, discusses the classification of SSCs.

Title 10 of the Code of Federal Regulations, Part 50.2, Definitions, provides definitions including that for safety-related SSCs. The definition states:

Safety-related structures, systems and components means those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or Page 4 of 28

(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in

§ 50.34(a)(1) or § 100.11 of this chapter, as applicable.

Title 10 of the Code of Federal Regulations, Part 70.4, Definitions, provides the definition for items relied on for safety (IROFS). The definition states:

Items relied on for safety mean structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the performance requirements in § 70.61 or to mitigate their potential consequences. This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e., beyond those in the minimum set necessary for compliance with the performance requirements) as items relied on for safety.

Title 10 of the Code of Federal Regulations, Part 21.3, Definitions, defines a basic component. As it would apply to the SHINE facility, the definition states:

basic component means a structure, system, or component, or part thereof, that affects their safety function, that is directly procured by the licensee of a facility or activity subject to the regulations in this part and in which a defect or failure to comply with any applicable regulation in this chapter, order, or license issued by the Commission could create a substantial safety hazard.

SHINE PSAR, Section 3.5.1.1, Nuclear Safety Classifications for SSCs, states:

SHINE uses a modified definition from 10 CFR 50.2 Definitions to develop the definition of SR [safety-related] SSCs, where appropriate, and utilizes a portion of 10 CFR 70.4 Definitions for the definition of IROFS SSCs.

In the responses to RAIs 3.5-1 and 3.5-4, SHINE proposes a six-part definition of safety-related structures, systems and components, which modifies the 10 CFR 50.2 definition of safety-related structures, systems, and components to include performance requirements for SSCs important to safety in the SHINE radioisotope production facility, and eliminates the classification of certain SSCs as IROFS. While five of the six parts of this modified definition are performance-based, the fourth part of this definition (i.e., [t]hat the potential for an inadvertent criticality accident is not credible) is not performance-based.

Additional information is needed for the NRC staff to evaluate the adequacy of SHINEs modified definition of safety-related structures, systems, and components. Furthermore, additional information is needed for NRC staff to determine the relationship between definition of basic component in 10 CFR 21.3 to SHINEs proposed definition of safety-related structures, systems, and components.

Additionally, the NRC staff notes that despite SHINEs statement that this revised definition of safety-related structures, systems, and components eliminates the classification of certain structures, systems, and components as items relied on for safety, the term IROFS still appears in several places in the SHINE PSAR.

Page 5 of 28

a. Provide a performance-based definition for the fourth part (i.e., with respect to the credibility of an inadvertent criticality accident) of the six-part definition of safety-related structures, systems, and components or provide a discussion of why it is not necessary.

b. Discuss how SHINEs definition of safety-related structures, systems, and components relates to the 10 CFR 21.3 definition of basic component. (i.e., Describe whether SHINEs defined safety-related structures systems, and components can also be considered basic components under 10 CFR 21.3).

c. Remove references to items relied of for safety from the SHINE PSAR if there are no longer SSCs that will bear this designation. Otherwise, clarify how IROFS will be utilized as part of SHINEs ongoing safety basis SHINE Response

a. The SHINE definition of safety-related SSCs was provided in Part a of the SHINE Response to RAI 3.5-1 (Reference 4) regarding implementation of the requirements of 10 CFR 70.61(d).

In order to ensure the SHINE definition of safety-related is performance-based and readily applicable to SSCs, SHINE has revised Item 4 of the SHINE definition of safety-related SSCs regarding implementation of the requirements of 10 CFR 70.61(d). The SHINE definition of safety-related SSCs has been revised as follows:

Safety-Related SSCs: Those SSCs that are relied upon to remain functional during normal conditions and during and following design basis events to assure:

1. The integrity of the primary system boundary;

2. The capability to shutdown the target solution vessel (TSV) and maintain the target solution in a safe shutdown (SSD) condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential exposures comparable to the applicable guideline exposures set forth in 10 CFR 20;

4. That all nuclear processes are subcritical, including use of an approved margin of subcriticality;

5. That acute chemical exposures to an individual from licensed material or hazardous chemicals produced from licensed material could not lead to irreversible or other serious, long-lasting health effects to a worker or cause mild transient health effects to any individual located outside the owner controlled area; or

6. That an intake of 30 mg or greater of uranium in soluble form by any individual located outside the owner controlled area does not occur.

SHINE has revised the PSAR to incorporate the revised SHINE definition of safety-related SSCs. A mark-up of the PSAR changes is provided in Attachment 1.

The non-public (proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 2. The public (non-proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 3.

b. SHINE considers safety-related SSCs (i.e., as defined in Part a of this response, above) to be basic components, as defined in 10 CFR 21.3.

Page 6 of 28

c. There are no longer any SSCs at the SHINE facility that bear the classification of items relied on for safety (IROFS). Subsection 1.3.3 of the PSAR inadvertently retains references to IROFS. Additionally, the Acronyms and Abbreviations list for Chapter 14 of the PSAR inadvertently retains a definition for the IROFS acronym, which is not used elsewhere in the chapter.

SHINE has revised the PSAR to remove the inadvertent references to IROFS. A mark-up of the PSAR changes is provided in Attachment 1. The non-public (proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 2. The public (non-proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 3.

Table 3.5b-1 of the PSAR contains multiple references to items relied on for safety, but only in the context of a direct quotation from 10 CFR 70.64. As applied to SHINE, these items are referred to as safety-related SSCs. Therefore, the references to items relied on for safety in Table 3.5b-1 will not be removed from the PSAR.

RAI 3.5-7 As required by 10 CFR 50.34(a)(4), the information in the PSAR shall contain [a] preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

In response to RAI 3.5-4 SHINE stated, in part, that the Control Room is part of a safety-related, Seismic Category I structure, but that the other SSCs are not defined as safety-related based on SHINEs definition of safety-related structures, systems, and components.

While SHINEs response to RAI 3.5-4 also states that safety-related systems are automatic and would put the facility in a safe condition without operator intervention, there is no discussion on how control room operators or other facility personnel will determine the facility is in a safe condition or how personnel will maintain the facility in a safe condition.

Under the conditions of a postulated design basis earthquake and a loss of offsite power, conditions could exist which would inhibit facility personnel from determining that the facility is in a safe condition and maintaining the facility in a safe condition using necessary safety-related equipment:

• SHINE PSAR Tables 2.3-2 and 2.3-3 show that the outdoor temperature can vary between 37°F and 104°F. While Control Room ventilation is supplied via Facility Ventilation Zone 4 (FVZ4), which is nonsafety-related, the Control Room has no heating, ventilation, and air conditioning (HVAC) system, which could impact Control Room habitability and equipment operability.

• The SHINE facility does not include emergency lighting, which could impact the ability of facility personnel to assess the facility status and to staff the Control Room.

• Since the Stack Release Monitoring System is not defined as safety-related, its unavailability could impact the ability of facility personnel to determine that there are no releases going up the stack.

Page 7 of 28

• Since the Health Physics Monitors are not defined as safety-related, their unavailability could impact the ability of facility personnel to assess levels of contamination during egress from the facility.

• Since the Facility Data and Communication System is not defined as safety-related, its unavailability could impact the ability of control room personnel to determine the facility status and communicate with other facility personnel and offsite agencies.

• In the event of an earthquake, there is the possibility for both onsite and offsite toxic releases and smoke from fire. Since the Facility Breathing Air System is not defined as safety-related, its unavailability could impact the ability of facility personnel to determine the facility status.

Additional information is needed on the design of the SHINE control room and other SSCs for NRC staff to determine the adequacy of the design for the prevention of accidents and the mitigation of the consequences of accidents.

Taking into account the conditions described above, provide additional information describing how the design of the SHINE control room and other SSCs will allow control room operators or other facility personnel to determine the facility is in a safe condition or how personnel will maintain the facility in a safe condition in the event of a postulated design basis earthquake with a loss of off-site power.

SHINE Response SHINE will provide a response to RAI 3.5-7 by April 24, 2015.

Section 3.5b - Radioisotope Production Facility RAI 3.5b-2 As required by 10 CFR 50.34(a)(4), an applicant needs to submit [a] preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

In response to RAI 3.5b-1, SHINE provided the information on the systems supported by the standby diesel generator, and stated that Criterion 7 of 10 CFR 70.64 is met by the Uninterruptible Power Supply Chain.

Additional information is needed for the NRC staff to evaluate the standby diesel generators ability to meet Criterion 7 of 10 CFR 70.64 to determine the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

Page 8 of 28

Provide additional information stating how the standby diesel generator meets Criterion 7 of 10 CFR 70.64, given that it and the systems it powers are classified as non-safety-related and how the Uninterruptible Power Supply Chain meets Criterion 7 of 10 CFR 70.64 during a loss of normal AC power.

SHINE Response Table 3.5b-1 of the PSAR provides SHINEs general evaluation of the design bases of the radioisotope production facility (RPF) against the baseline design criteria specified under 10 CFR 70.64 and general design criteria (GDC) 61, 62, 63, and 64 of Appendix A to 10 CFR 50 as a good design practice.

Criterion 7, Utility services, of the baseline design criteria specified under 10 CFR 70.64 states, The design must provide for continued operation of essential utility services.

While the standby diesel generator (SDG) provides an additional source of power should off-site power be lost, it is not safety-related and is not used to meet Criterion 7 of 10 CFR 70.64.

SHINE plans to use the uninterruptible electrical power supply system (UPSS) to power safety-related equipment should off-site power be lost and the SDG not be available. The UPSS provides continued operation of essential electrical utility service at the SHINE facility, and no other essential utilities are identified for the facility.

SHINE will revise the As Applied and Means of Compliance discussion provided for Criterion 7 in Table 3.5b-1 in the FSAR as shown below. The revised discussion is consistent with the As Applied and Means of Compliance discussion provided for Criterion 17, Electric power systems, in Table 3.5a-1 of the PSAR. An IMR has been initiated to track the revision to Table 3.5b-1 in the FSAR.

BASELINE DESIGN CRITERIA 10 CFR 70.64 As Applied to SHINE (7) Utility services. The design must provide As Applied and Means of Compliance for continued operation of essential utility services. An on-site electric power system (defined to be the SHINE facility provided power supply) and an off-site (defined to be the public utility transmission network supplied substations) electric power system are provided to permit functioning of structures, systems, and components important to safety.

The functions for each system (assuming the other system is not functioning) are to provide sufficient capacity and capability to ensure that the RPF processes remain in a safe condition and confinement integrity and other vital functions are maintained in the event of postulated accidents.

The on-site electric power supply systems include a safety-related UPSS, a commercial grade diesel generator system, and the on-site electric distribution system.

The UPSS has sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. Each of the redundant UPSS circuits are designed to be available within a few seconds following a loss of all off-site electric power, to ensure that safety systems requiring Page 9 of 28

electrical power (such as RAMS and CAAS) can perform their safety functions.

The diesel generator system performs no safety functions but provides an additional degree of defense-in-depth to the electric power system.

Portions of the on-site electric distribution, those circuits which operate in conjunction with the UPSS, are safety-related.

The SHINE facility receives a single physically off-site independent power circuit. This off-site power circuit consists of two power feeds connected to two local outdoor transformers.

Provisions are included to minimize the probability of losing electric power from the UPSS as a result of or coincident with, the loss of power from the transmission network.

Refer to Section 6b and Chapter 8 for a detailed discussion.

Page 10 of 28

CHAPTER 6 - ENGINEERED SAFETY FEATURES Section 6b.3 - Nuclear Criticality Control (Applies to RAIs 6b.3-23 through 30)

As required by 10 CFR 50.34(a)(4), an applicant needs to submit [a] preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

As stated in the ISG Augmenting NUREG-1537, Chapter 13, the NRC staff has determined that the use of integrated safety analysis (ISA) methodologies as described in 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, and NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, Revision 1, May 2010, application of the radiological and chemical consequence and likelihood criteria contained in the performance requirements of 10 CFR Section 70.61, designation of IROFS, and establishment of management measures are acceptable ways of demonstrating adequate safety for the medical isotopes production facility.

Applicants may propose alternate accident analysis methodologies, alternate radiological and chemical consequence and likelihood criteria, alternate safety features, and alternate methods of assuring the availability and reliability of the safety features.

As used in this ISG, the term performance requirements, when referencing 10 CFR Part 70, Subpart H, is not intended to mean that the performance requirements of Subpart H are required for a radioisotope production facility license, only that their use as accident consequence and likelihood criteria may be found acceptable by NRC staff.

RAI 6b.3-23 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, Nuclear Criticality Safety for the Processing Facility, states, in part that [c]riticality process safety controls should be provided for criticality safety, and a description of their safety function should be described. The applicant should use enough safety controls to demonstrate that, under normal and abnormal credible conditions, all nuclear processes remain subcritical and that NCS [nuclear criticality safety]

limits on controlled parameters will be established to ensure that all nuclear processes are subcritical, including an adequate margin of subcriticality for safety.

For example, the applicant could commit to base the safety limits on validated calculation methods. These methods should be industry-accepted and peer-reviewed. Also, the applicant should commit to ensuring that methods used to develop NCS limits will be validated to confirm that they are used within acceptable ranges and that the applicant used both appropriate assumptions and acceptable computer codes.

In response to RAI 6b.3-1b, SHINE submitted a general validation report and a project-specific validation report, which provided the methods and assumptions used to determine that nuclear criticality safety criteria are met at the SHINE facility. Staff reviewed these reports and has Page 11 of 28

determined that additional information is needed to determine that the methods used to validate NCS safety criteria are acceptable and that SHINE used both appropriate assumptions and acceptable computer codes.

a. The general validation report appears to be an off-the-shelf report, dated 2007. As described in ANSI/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, verification and validation should be performed using the same operating systems, software, and hardware that will be used for performing evaluations and placing systems under configuration control.

Provide additional information describing how the general validation report uses appropriate methods to validate NCS safety criteria, including information on the operating systems, software, and hardware used to perform evaluations and place systems under configuration control.

b. As described in NUREG-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, prior to the initiation of validation activities, the operating conditions and parameters for which the validation is to apply must first be identified.

For both the general and project-specific validation reports, describe the area of applicability with respect to the actual operations and describe the applicability of the benchmark experiments to these operations.

c. As described in NUREG-6698, the statistical results from the bias trends are used to establish safety limits. Both the general and project-specific validation reports are missing an evaluation of trends in the bias data, which may impact potential bias estimates.

For both the general and project-specific validation reports, provide an evaluation of the trends in the bias data, describing potential impacts on the bias impacts.

d. The validation of two different libraries for different materials, introduces the possibility of human error in selecting a library for the evaluation (i.e., picking the wrong library).

Provide the methods SHINE uses to guard against the selection of the incorrect library for validation.

e. The project-specific validation report utilizes only a limited number of experiments to evaluate the bias and estimate uncertainty. For example, only four experiments were listed for the most applicable enrichment. Additionally, modeled results were compared with calculated results, as opposed to only with experimental data.

Explain why only a limited number of experiments are sufficient to evaluate bias and estimate uncertainty. Additionally, explain why modeled results were compared with calculated results, as opposed to only with experimental data.

SHINE Response SHINE will provide a response to RAI 6b.3-23 by April 24, 2015.

Page 12 of 28

RAI 6b.3-24 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, states, in part, that [e]ach process that has accident sequences leading to criticality should have sufficient controls in place to ensure double-contingency protection. This may be provided by either (1) control of two independent process parameters, or (2) control of a single process parameter, such that at least two independent failures would have to occur before criticality is possible. The first method is preferable because of the inherent difficulty in preventing common-mode failure when controlling only one parameter. Additional discussions of the double-contingency principle are available on pages 5-7 and 5-A-8 of NUREG-1527.

In response to RAI 6b.3-10, SHINE states that tank 1-TSPS-01T meets double contingency by geometry and the configuration management program. NRC staff does not view the configuration management program as an independent control. As described in 10 CFR 70.4, configuration management is not normally considered a management measure because it assures the availability of the geometry control.

Provide additional information describing how tank 1-TSPS-01T meets the double contingency principle.

SHINE Response SHINE will provide a response to RAI 6b.3-24 by April 24, 2015.

RAI 6b.3-25 NUREG 1520, Section 3.4.3.2, Integrated Safety Analysis Summary and Documentation, states that an event defined as not credible must be convincing despite the absence of designated [controls]. Typically, this can be achieved only for external events known to be extremely unlikely.

In response to RAI 6b.3-3, SHINE provides the basis for considering a criticality sequence to be not credible. In this explanation, SHINE essentially quotes the three independent acceptable sets of qualities that could define an event as not credible from Section 3.4.3.2 of NUREG-1520.

In the response to RAI 13b.1-1, SHINE established, in Table 13b.1-1-2, Likelihood Index Limit Guidelines, an event frequency associated with highly unlikely and unlikely that is consistent with the values found in NUREG-1520.

While SHINE provides three independent sets of qualities that could define an event as not credible, SHINEs criteria does not take into account an absence of designated controls.

Without a reliance on controls, consistent with the guidance in NUREG-1520, criticality, in certain situations, would be credible, but highly unlikely, due to the establishment of controls consistent with the methods proposed. Therefore, additional information is required for NRC staff to determine the adequacy of SHINEs use of the term not credible in its PSAR with respect to criticality safety.

Page 13 of 28

Revise the definition of not credible to not include reliance on controls consistent with the guidance in NUREG-1520 and event frequency limits provided in Table 13b.1-1-2 or provide information why this is not necessary.

SHINE Response SHINE will provide a response to RAI 6b.3-25 by April 24, 2015.

RAI 6b.3-26 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, states, in part, that the reviewer should determine whether the margin of subcriticality for safety is sufficient to provide reasonable assurance of subcriticality.

In response to RAI 6b.3-4, SHINE states it intends to utilize a subcritical margin of 0.05 with additional considerations for uncertainty in the validation and modeling. In addition, SHINE states in multiple places in the PSAR that processes will be maintained to a keff 0.95 (assuming a subcritical margin of 0.05).

The NRC staffs review of SHINEs response to RAI 6b.3-1, which requested the applicants validation report and NCS reference manual, found that there was insufficient benchmarking of the code against experiments utilizing the materials and enrichments expected in SHINEs processes. For this reason, the proposed subcritical margin of 0.05 is not sufficient to adequately address the uncertainty associated with the neutron interactions of these process materials. The subcritical margin of 0.05, which SHINE quoted from NUREG-1520, was intended for facilities with enrichment less than five percent utilizing well established processes and for which there is significant experience and data. In contrast, the SHINE facility will be a first-of-a-kind facility using materials not normally utilized and of an enrichment up to 20 percent.

Provide additional information describing how SHINE will sufficiently benchmark against experiments utilizing the materials and enrichments expected to be used in SHINE facility processes for its proposed margin of subcriticality, or propose a new margin of subcriticality that appropriately takes into account materials and enrichment.

SHINE Response SHINE will provide a response to RAI 6b.3-26 by April 24, 2015.

RAI 6b.3-27 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, Nuclear Criticality Safety for the Processing Facility, states that the reviewer should determine if, when they are relevant, the applicant considers heterogeneous effects. Heterogeneous effects are particularly relevant for low-enriched uranium processes, where, all other parameters being equal, heterogeneous systems are more reactive than homogeneous systems.

SHINE PSAR, Section 6b.3, Nuclear Criticality Control, states that [h]eterogeneous effects are not considered applicable because the uranium enrichment is less than 20 percent.

Page 14 of 28

In the response to RAI 6b.3-6, SHINE elaborated on its treatment of heterogeneous effects by quoting LA-12808, Nuclear Criticality Safety Guide, stating that heterogeneous effects can be ignored for uranium with an enrichment above six percent uranium-235.

As shown in figures 22 through 25 of LA-10860-MS, Critical Dimensions of Systems Containing U-235, Pu-239, and U-233, heterogeneity does affect some parameters at greater than six percent enrichment in U-235.

Provide additional information acknowledging that heterogeneity effects will be considered when establishing NCS controls and limits, where such are credible and relevant.

SHINE Response SHINE will provide a response to RAI 6b.3-27 by April 24, 2015.

RAI 6b.3-28 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, Nuclear Criticality Safety for the Processing Facility, states that the reviewer should determine whether the applicants use of moderator as a controlled parameter is acceptable.

In the response to RAI 6b.3-8, SHINE states that preliminary criticality scoping safety assessments include optimum moderation conditions. The preliminary design does not contain systems that require moderation as the sole controlled criticality safety parameter and there are no plans to have moderation controlled areas. While moderation may not be a sole controlled safety parameter, it is unclear whether SHINE will be reliant on moderation controls in any capacity. NRC staff needs additional information to determine whether SHINE considers moderation to be a controlled parameter.

Verify whether the SHINE facility will be reliant on moderation controls in any capacity as opposed to a sole control. If moderation controls will be used, provide information describing these controls.

SHINE Response SHINE will provide a response to RAI 6b.3-28 by April 24, 2015.

RAI 6b.3-29 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, Nuclear Criticality Safety for the Processing Facility, states that the reviewer should determine whether the applicants use of enrichment as a controlled parameter is acceptable.

In response to RAI 6b.3-18 SHINE states that criticality scoping safety assessments have been performed using a uranium enrichment value of 21 percent to conservatively address uranium enrichment. NRC staff needs additional information to determine whether SHINE considers enrichment to be a controlled parameter.

Page 15 of 28

Verify whether enrichment will be a controlled parameter controlled by independent verification, upon receipt of materials, to assure no out-of-spec materials are utilized.

SHINE Response SHINE will provide a response to RAI 6b.3-29 by April 24, 2015.

RAI 6b.3-30 The ISG Augmenting NUREG-1537, Part 2, Section 6b.3, states, in part, that [c]riticality process safety controls should be provided for criticality safety, and a description of their safety function should be described. The applicant should use enough safety controls to demonstrate that, under normal and abnormal credible conditions, all nuclear processes remain subcritical.

While SHINE states in response to RAIs 6b.3-1 and 6b.3-22, that the NCS reference manual and formal NCSEs have yet to be generated, the NRC staff need additional information to determine that enough safety controls have been considered to demonstrate that, under normal and abnormal credible conditions, all nuclear processes remain subcritical.

Provide additional information discussing the methodologies and assumptions that will be used to develop SHINEs NCS reference manual and provide a representative sample of nuclear criticality safety evaluations to demonstrate the methods used to demonstrate that under normal and abnormal credible conditions, all nuclear processes remain subcritical.

SHINE Response SHINE will provide a response to RAI 6b.3-30 by April 24, 2015.

Page 16 of 28

CHAPTER 12 - CONDUCT OF OPERATIONS RAI 12.7-36 NUREG-0849, Standard Review Plan for the Review and Evaluation of Emergency Plans for Research and Test Reactors, Section 3.0, Organization and Responsibilities, Evaluation Item 1.b., states that the emergency plan should describe [t]he reactor's emergency organization, including augmentation of the reactor staff to provide assistance for coping with the emergency situation, recovery from the emergency, and maintaining emergency preparedness.

On September 19, 2014, NRC staff requested additional information on the actions to be taken by operators when an emergency is declared. Specifically, in RAI 12.7-4, SHINE was asked to describe the actions the on-shift operators will take if they cannot ensure their activities can be placed in a safe condition before reporting to the on-site assembly area.

SHINEs responses to RAIs 12.7-4 and 12.7-35 did not adequately address the actions of the on-shift operators, such as the Shift Supervisor, Senior Facility Operator, and Facility Operator, when an emergency is declared.

In order for the NRC staff to determine the adequacy of SHINEs emergency organization, additional information is needed on the actions of the on-shift operators when an emergency is declared.

a. In the event of an emergency, describe the actions that will be taken by on-shift operators if they cannot ensure their activities can be placed in a safe condition.

b. Following a design-basis, beyond design-basis, security, unplanned (e.g., a radiological release or chemical spill), or other type of on-site or off-site event, describe the actions taken by on-shift operators to ensure their activities are placed or remain in a safe condition.

c. In the event of an emergency, describe the programmatic process(es) for individuals and their activities, as well as the safety systems necessary to maintain safe shutdown.

SHINE Response

a. SHINE will describe the actions that will be taken by on-shift operators if they cannot ensure their activities can be placed in a safe condition in the event of an emergency in the SHINE Emergency Plan, which will be provided as part of the SHINE Operating License (OL)

Application. An IMR has been initiated to ensure a description of the actions that will be taken by on-shift operators if they cannot ensure their activities can be placed in a safe condition in the event of an emergency is provided in the SHINE Emergency Plan.

b. SHINE will describe the actions that will be taken by on-shift operators to ensure their activities are placed or remain in a safe condition following a design-basis, beyond design-basis, security, unplanned (e.g., a radiological release or chemical spill), or other type of on-site or off-site event in the SHINE Emergency Plan, which will be provided as part of the SHINE OL Application. An IMR has been initiated to ensure a description of the actions that will be taken by on-shift operators to ensure their activities are placed or remain in a safe condition following a design-basis, beyond design-basis, security, unplanned (e.g., a radiological release or chemical spill), or other type of on-site or off-site event is provided in the SHINE Emergency Plan.

Page 17 of 28

c. SHINE will describe the programmatic processes for individuals and their activities, as well as the safety systems necessary to maintain safe shutdown in the event of an emergency, in the SHINE Emergency Plan, which will be provided as part of the SHINE OL Application.

An IMR has been initiated to ensure a description of the programmatic processes for individuals and their activities, as well as the safety systems necessary to maintain safe shutdown in the event of an emergency, is provided in the SHINE Emergency Plan.

Page 18 of 28

APPENDIX 12C - QUALITY ASSURANCE PROGRAM DESCRIPTION Appendix 12C Section 1 - Introduction RAI 12C.1-5 As required by 10 CFR 50.34(a)(7), each applicant for a construction permit to build a production or utilization facility must include, in its preliminary safety analysis report, a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. Regulatory Guide 2.5, Revision 1, states that the general requirements for establishing and executing a quality assurance program for the design, construction, testing, modification, and maintenance of research and test reactors in the American National Standards Institute/American Nuclear Society Standard (ANSI/ANS) 15.8-1995 provide an acceptable method for complying with the program requirements of 10 CFR 50.34, Contents of applications; technical information.

Section 12.9, Quality Assurance, of the SHINE PSAR states that the SHINE QA-1, Quality Assurance Program Description (QAPD), is based on ANSI/ANS 15.8-1995 (R2005)

(ANSI/ANS, 1995), Quality Assurance Program Requirements for Research Reactors, with guidance from Regulatory Guide 2.5, Revision 1. However, it is not clear to what extent ANSI/ANS 15.8-1995 has been applied to the development of the SHINE QAPD for the facility.

SHINEs response to RAI 12C.1-3 stated, in part, SHINE has revised the Executive Summary of the SHINE Quality Assurance Program Description (QAPD) to state, SHINE has determined that ANSI/ANS 15.8-1995 is appropriate for use in the design of the facility even though this standard was written for research reactors.

In reference to SHINEs response to RAI 12C.1-3, ANSI/ANS 15.8-1995 is not a design standard and cannot be directly applied to the design of the SHINE facility.

Clarify whether SHINE has determined that ANSI/ANS 15.8-1995 is sufficient for use in the development of the SHINE QAPD, which is to be applied in the design, fabrication, construction, and operation of the SHINE facility.

SHINE Response SHINE has revised the Executive Summary of the SHINE Quality Assurance Program Description (QAPD) to state, SHINE has determined that ANSI/ANS-15.8-1995 is sufficient for use in the development of the SHINE QAPD, which is to be applied to the design, fabrication, construction, and operation of the SHINE facility. Revision 6 of the SHINE QAPD is provided as Attachment 2.

RAI 12C.1-6 ANSI/ANS-15.8-1995, Section 1.3, Definitions, defines an experiment as [a]ny operation, hardware, or target (excluding devices such as detectors, foils, etc.) that is designed to investigate non-routine reactor characteristics or that is intended for irradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidly secured to a core or shield structure so as to be a part of their design.

Page 19 of 28

The SHINE QAPD, Section 2.10, Inspections, states, in part, that [t]he inspection program shall apply to procurement, construction, modification, maintenance, and experiment fabrication.

The SHINE QAPD Section 2.19, Experimental Equipment, states that [t]he quality assurance program shall provide controls over the design, fabrication, installation and modification of experimental equipment to the extent that these impact safety-related items.

The SHINE QAPD, Section 1.3, Definitions, does not provide a definition of an experiment.

Further, SHINEs response to RAI 12C.1-4 stated, in part, that The definition of experiment is not listed in the SHINE QAPD, as SHINE does not plan on conducting experiments as defined in ANSI/ANS-15.8-1994 (R2013).

Addtionally, it is not clear how SHINEs definitions of audit and assessment are related in Sections 2.7, Control of Purchased Items and Services, and Section 2.18, Assessment.

a. Clarify the basis for not providing a definition of experiment, if SHINE is planning to conduct experiments and utilize experimental equipment, as discussed in Sections 2.10 and 2.19 of the SHINE QAPD, respectively.

b. Provide additional information regarding the definitions of audit, as used in the SHINE QAPD Sections 2.7, Control of Purchased Items and Services, and 2.7.3, Verification Activities, and assessment, as used in Section 2.18, Assessment. Provide clarification as to the difference between the two SHINE Response

a. SHINE does not plan on conducting experiments or utilizing experimental equipment; therefore, a definition of experiment is not provided in the SHINE QAPD. SHINE has revised Section 2.10 of the SHINE QAPD to remove the phrase experiment fabrication, and has removed Section 2.19, Experimental Equipment, from the SHINE QAPD.

Revision 6 of the SHINE QAPD is provided as Attachment 2.

b. The SHINE definition of assessment and audit are identical, and SHINE uses the terms interchangeably. SHINE defines both assessment and audit as:

A planned and documented activity performed to determine by investigation, examination, or evaluation of objective evidence the adequacy of and compliance with established procedures, instructions, drawings, and other applicable documents, and the effectiveness of implementation. An audit/assessment should not be confused with surveillance or inspection activities performed for the sole purpose of process control or product acceptance.

Based on the above, there is no difference between the SHINE definition of assessment and audit.

Page 20 of 28

Appendix 12C Enclosure 2 - Graded Approach to Quality RAI 12C.E2-6 ANSI/ANS-15.8-1995, Section 2.2, Quality Assurance Program, states, in part, that [t]he program shall identify the items and activities to which it applies and the extent of program application for each item and activity.

The SHINE QAPD Enclosure 2 - Graded Approach to Quality, states, in part, that [t]he graded approach to quality is a process by which the level of analysis, documentation, and actions necessary to comply with a requirement is commensurate with the safety significance. It further states that QL-1 shall implement the full measure of this QAPD and shall be applied to Safety-Related Structures, Systems and Components. QL-2 will include the non-safety related quality activities performed by the licensee, that are deemed necessary by SHINE to ensure the manufacture and delivery of highly reliable products and services to meet or exceed customer expectations and requirements.

SHINEs response to RAI 12C.E2-5 stated, in part, that [t]he QL-2 classification is applicable to all nonsafety-related SSCs and activities, and is applied by SHINE to ensure the manufacture and delivery of highly reliable products and services to meet or exceed customer expectations and requirements. It further stated that SHINE design control and procurement procedures will consider the function of the nonsafety-related SSC or activity, the relative importance of that function, whether there are any codes, standards, or other requirements that apply, and determine the appropriate quality measures.

Addtionally, the definition of the QL-2 classification provided in the SHINE QAPD focuses on ensuring the manufacture and delivery of highly reliable products and services to meet or exceed customer expectations and requirements. Furthermore, the definition of the QL-2 classification provided in the QAPD does not indicate that the QL-2 classification is applicable to all non-safety related SSCs and activities, whereas and SHINEs response to RAI 12C.E2-5 states that it does.

a. Clarify if the QL-1 classification applies to safety-related activities, as well as safety-related SSCs.

b. Clarify how the definition of the QL-2 classification provided in the SHINE QAPD is based on safety significance considerations. Further, clarify whether the definition of the QL-2 classification is based on the application of the full scope of the QAPD to the activities affecting quality, as opposed to only design and procurement requirements, as described in the response to RAI 12C.E2-5.

c. Clarify if the QL-2 classification is intended to be applied only to selected non-safety related SSCs and activities.

Page 21 of 28

SHINE Response

a. The revised SHINE definition of the QL-1 classification applies to safety-related activities, as well as safety-related SSCs. SHINE has revised the definition of the QL-1 classification in Enclosure 2 of the SHINE QAPD and Subsection 3.5.1.2.1 of the PSAR to read as follows:

QL-1 shall implement the full measure of this QAPD and shall be applied to safety-related structures, systems, and components and to safety-related activities.

Revision 6 of the SHINE QAPD is provided as Attachment 2. A mark-up of the PSAR changes is provided in Attachment 1. The non-public (proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 2. The public (non-proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 3.

b. The current SHINE definition of the QL-2 classification in the SHINE QAPD is not based on safety significance considerations. SHINE has revised the definition of the QL-2 classification in Enclosure 2 of the SHINE QAPD and Subsection 3.5.1.2.2 of the PSAR to now be based on safety significance considerations. In addition, the revised definition of the QL-2 classification is based on the application of the full scope of the SHINE QAPD to the activities affecting quality, as opposed to only design and procurement requirements, as described in the SHINE Response to RAI 12C.E2-5 (Reference 5). SHINE has revised the definition of the QL-2 classification to read as follows:

QL-2 is applied to selected structures, systems, and components and activities intended to support or protect the safety function of safety-related equipment. Quality Assurance Program elements are applied to an extent that is commensurate with the items importance to safety. Implementing documents establish program element applicability.

In addition, SHINE has added a new QL-3 classification that addresses SSCs and activities that may be important to ensuring customer requirements are met, or operational or mission-related goals such as throughput, cost, or schedule are achieved. Controls, appropriate for the application, are applied to SSCs and activities using the SHINE Quality Assurance Program for efficiency to avoid the creation and use of a separate or redundant management system. The new definition of the QL-3 classification reads as follows:

QL-3 is applied to nonsafety-related structures, systems, and components and activities and does not support or protect the safety function of safety-related structures, systems, and components or activities. However, the performance of QL-3 structures, systems, and components and activities may be important to ensuring customer requirements are met, or operational or mission-related goals such as throughput, cost, or schedule are achieved. Controls, appropriate for the application, are applied to structures, systems, and components and activities using the SHINE Quality Assurance Program for efficiency to avoid the creation and use of a separate or redundant management system.

The controls on these structures, systems, and components and activities do not impact QL-1 or QL-2 structures, systems, and components and activities or the regulatory basis of the facility.

Page 22 of 28

The quality groups provided in Table 3.5-1 of the PSAR have been revised to reflect the revision to the definition of the QL-2 classification and the new definition of the QL-3 classification.

Revision 6 of the SHINE QAPD is provided as Attachment 2. A mark-up of the PSAR changes is provided in Attachment 1. The non-public (proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 2. The public (non-proprietary) version of the PSAR, incorporating the changes provided in Attachment 1, is provided in Enclosure 3.

c. The SHINE QL-2 classification is intended to be applied only to selected nonsafety-related SSCs and activities.

Page 23 of 28

CHAPTER 13 - ACCIDENT ANALYSIS Section 13a2.2 - Accident Analysis and Determination of Consequences (Applies to RAIs 13a2.2-5 through 7)

The ISG Augmenting NUREG-1537, Part 2, Section 13a2, Aqueous Homogeneous Reactor Accident Analyses, states that the applicant should include a systematic analysis and discussion of credible accidents for determining the limiting event in each category and that the mathematical models and analytical methods employed, including assumptions, approximations, validation, and uncertainties, should be clearly stated.

RAI 13a2.2-5 While SHINEs response to RAI 13a2.2-1 states that the basis for the internal dose conversion factors (DCFs) is International Commission on Radiological Protection Publication 30, Limits for Intakes of Radionuclides by Workers, (ICRP 30) additional information is needed on this basis, as applied to offsite DCFs for members of the public, for the NRC staff to determine the adequacy of SHINEs radiological dose consequence analysis as part of its accident analysis.

Provide information supporting the acceptability of ICRP 30, which provides DCFs for adult workers, as the basis for calculating offsite doses to members of the public, including children.

SHINE Response SHINE will provide a response to RAI 13a2.2-5 by April 24, 2015.

RAI 13a2.2-6 In response to RAI 13a2.2-1, SHINE conservatively uses a respirable fraction (RF) of 1.0 for most accidents listed in Table 13a2.2.1-4, including the Mishandling and Malfunction of Target Solution. However, a value of 0.4 is used for the particulate RF in the maximum hypothetical accident, which is similar to the Mishandling and Malfunction of Target Solution. Additional information is needed for the NRC staff to determine the adequacy of SHINEs radiological dose consequence analysis as part of its accident analysis.

Provide a basis for the use of a RF value of 0.4 used in the maximum hypothetical accident, including an explanation for why it is different than the RF for the Mishandling and Malfunction of Target Solution accident.

SHINE Response The Respirable Fraction (RF) of particulates is dependent on the droplet size of the airborne material. The postulated event for the IF postulated maximum hypothetical accident (MHA) is a large rupture of the TSV and Subcritical Assembly Support Structure (SASS) resulting in a complete release of the target solution and fission product inventory into one IU cell. As described in the SHINE Response to RAI 13a2.1-8 (Reference 6), the presence of the light water pool is ignored for the purposes of the IF postulated MHA.

Page 24 of 28

With the TSV maintained slightly below atmospheric pressure and the SASS operating near atmospheric pressure, the scenario from NUREG/CR-6410 (Reference 7) that was determined to be most applicable was a free-fall spill of aqueous liquid. The Airborne Release Fraction (ARF) and RF that are listed for this scenario, 2E-4 and 0.4 respectively, represent the maximum ARF x RF values in the experimental data set that the values were derived from.

Alternative values for RFs in the same experimental data set reach higher values; however, the combined ARF x RF values are not bounding.

The postulated event for the mishandling and malfunction of target solution is a TSV dump tank piping rupture into the IU cell. The scenario from NUREG/CR-6410 of a pressurized vessel being vented below the surface level of the liquid was chosen as a conservative approach.

NUREG/CR-6410 lists a combined ARF x RF value of 1E-4 for this scenario, which is chosen based on sprays generated by commercial spray nozzles designed to reduce droplet sizes. The pressure generated in SHINEs transfer of solution from the dump tank is expected to be relatively low (e.g., less than 100 psig), and as stated in DOE-HDBK-3010-94 (the referenced source for the event in NUREG/CR-6410) (Reference 8), it is not anticipated that drops formed from breaches, cracks, or leaks would generate finer drop size distributions than the commercial spray nozzles used in the study. Therefore, the chosen ARF x RF values conservatively bound the postulated event.

In summary, the RF values for these two accident scenarios differ because of the method in which the material was assumed to escape the vessel. Both sets of ARF x RF values were chosen from the events in NUREG/CR-6410 judged most appropriate for the specific scenario under consideration and the values are conservative.

RAI 13a2.2-7 While SHINEs response to RAI 13a2.2-4 indicates that MCNP5 was used to calculate fluxes and cross-sections specific to various target solution vessel configurations, additional information is needed for NRC staff to determine the adequacy of SHINEs radiological dose consequence analysis as part of its accident analysis, recognizing that the typical use of MCNP involves using existing cross section libraries.

Provide additional information on how MCNP is used to calculate cross sections and what these cross sections are used for at for the SHINE facility.

SHINE Response As described in the SHINE Response to RAI 13a2.2-4 (Reference 6), SHINE used the COUPLE and ORIGEN-S codes, part of the larger SCALE 6.1 computation system, to generate source term data.

COUPLE collapses the energy-dependent cross section data into one-group cross section data for use in the time-dependent differential equation set solved by ORIGEN-S. These one-group cross sections are no longer physical properties of the isotopes, but are problem-specific cross sections.

Page 25 of 28

COUPLE can collapse the cross sections itself using a known flux spectrum, or it will accept user-defined one-group cross sections determined from another source. SHINE used Monte Carlo N-Particle (MCNP) to calculate user-defined one-group cross sections by tallying the integral of the product of the neutron flux and the nuclear cross-section of interest and then dividing by the total flux. Since MCNP is a continuous energy neutronics program, more accurate cross-section collapsing is realized by this method rather than using group structure discretization.

These one-group cross sections are used in COUPLE and ORIGEN-S to generate the radionuclide inventory in the target solution following irradiation.

Section 13b.1 - Radioisotope Production Facility Accident Analysis Methodology RAI 13b.1-3 As required by 10 CFR 50.34(a)(4), the preliminary safety analysis report should include [a]

preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

The ISG Augmenting NUREG-1537, Part 2, Section 13b.1.2, Accident Initiating Events, states that [i]nformation in the [safety analysis report] should allow the reviewer to follow the sequence of events in the accident scenario from initiation to a stabilized condition. The reviewer should confirm the following:

• The credible accidents were categorized, and the most limiting accident in each group was chosen for detailed analyses.

• The process was assumed to be operating normally under applicable technical specifications before the initiating event. However, the process may be in the most limiting technical specification condition at the initiation of the event.

• Instruments, controls, and automatic protective systems were assumed to be operating normally or to be operable before the initiating event. Maximum acceptable non-conservative instrument error may be assumed to exist at accident initiation.

• The single malfunction that initiates the event was identified.

• Credit was taken during the scenario for normally operating process systems. Protective actions were initiated by either the operating staff, control systems, or ESFs.

• The sequence of events and the components and systems damaged during the accident scenario were clearly discussed.

• Validated mathematical models and analytical methods that were employed, including assumptions, approximations, and uncertainties, were clearly stated.

• The radiation source terms were presented or referenced.

Page 26 of 28

• The potential radiation consequences to the facility staff and the public were presented and compared with acceptance criteria.

While Tables 13b.1-2-1 and 13b.1-1-4 provided in response to RAIs 13b.1-1 and 13b.1-2 provided potential accident sequences at the SHINE facility, additional information is needed for the NRC staff to evaluate SHINEs methodology for determining the sequence of events in an accident scenario from initiation to a stabilized condition.

Provide detailed accident sequence descriptions for at least four of the sequences listed in Table 13b.1-1-4, from the initiating events through the sequences mitigated consequences. The descriptions should include the most limiting examples from a chemical accident, radiological accident, fire accident, and criticality accident, and should discuss the following:

a) The hazards involved in the accident, including source terms and initiating events. The initiating events should consider potential operator errors as well as external events and equipment failures; b) An explanation of the methodology for selecting the numerical value of the likelihood of the initiating event; c) An analysis of the unmitigated consequences, including the classification of unmitigated consequences as radiological or chemical, as well as less-than-intermediate, intermediate, or high; d) A description of the accident progression, and the function(s) of IROFS, as applicable, in the accident sequence mitigation; e) An explanation of methodology for selecting the numerical value for the availability and reliability of the IROFS, as applicable, and the resulting mitigated likelihood; and f) An explanation of how the mitigated consequences were estimated.

SHINE Response SHINE will provide a response to RAI 13b.1-3 by April 24, 2015.

Page 27 of 28

References (1) NRC letter to SHINE Medical Technologies, Inc., dated March 25, 2015, SHINE Medical Technologies, Inc. - Request for Additional Information Regarding Application for Construction Permit (TAC Nos. MF2305, MF2307, and MF2308) (ML15055A116)

(2) SHINE Medical Technologies, Inc. letter to NRC, dated March 26, 2013, Part One of the SHINE Medical Technologies, Inc. Application for Construction Permit (ML130880226)

(3) SHINE Medical Technologies, Inc. letter to NRC, dated May 31, 2013, Part Two of the SHINE Medical Technologies, Inc. Application for Construction Permit (ML13172A324)

(4) SHINE Medical Technologies, Inc. letter to NRC, dated October 15, 2014, SHINE Medical Technologies, Inc. Application for Construction Permit, Response to Request for Additional Information (ML14296A190)

(5) SHINE Medical Technologies, Inc. letter to NRC, dated February 6, 2015, SHINE Medical Technologies, Inc. Application for Construction Permit, Response to Request for Additional Information (ML15043A404)

(6) SHINE Medical Technologies, Inc. letter to NRC, dated December 3, 2014, SHINE Medical Technologies, Inc. Application for Construction Permit, Response to Request for Additional Information (ML14356A528)

(7) U.S. Nuclear Regulatory Commission, Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410, March 1998 (ML072000468)

(8) U.S. Department of Energy, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, DOE-HDBK-3010-94, December 1994 (Reaffirmed 2013)

Page 28 of 28