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University of Missouri - Columbia Research Reactor - Written Communication as Specified by 10 CFR 50.4(b)(l) Requesting U.S. Nuclear Regulatory Commission Approval to Amend the Technical Specifications
	ML17089A229
Person / Time
Site:	University of Missouri-Columbia
Issue date:	03/27/2017
From:	Meffert B A; Univ of Missouri - Columbia
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML17089A229 (233)
	v • d • e

EiJ University of Missouri March 27, 2017 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-37 Washington, DC 20555-0001 REFERENCE: Docket No. 50-186 University of Missouri -Columbia Research Reactor Renewed Facility Operating License No. R-103 Research Reactor Center 1513 Research Park Drive Columbia, MO 65211 PHONE 573-882-4211 WEB murr.missouri.edu

SUBJECT:

Written communication as specified by 10 C.F.R. 50.4(b)(l) requesting U.S. Nuclear Regulatory Commission approval to amend the Technical Specifications appended to Renewed Facility Operating License No. R-103 pursuant to 10 C.F.R. 50.59(c) and 10 C.F.R. 50.90 1. Introduction The University of Missouri Research Reactor (MURR) is requesting a license amendment and changes to the facility Technical Specifications (TSs) in order to conduct a more extensive experiment that will produce molybdenum-99 (Mo-99) in large quantities as part of its role in supplying critical medical radioisotopes to the domestic and international community. This experiment would utilize General Atomics' (GA) Selective Gas Extraction (SGE) process, which consists of irradiating target rods containing low-enriched uranium (LEU) pellets in the reactor graphite reflector region in order to produce fission product Mo-99. The Mo-99 would then be extracted from the LEU in hot cells using the SGE technology (which does not require chemical dissolution of the LEU targets, thus avoiding the creation of large quantities of mixed liquid radioactive waste).1 As described in detail below, this amendment request would only apply to the "in-pool" portion of the project, or Part 1, where the LEU target rods would be irradiated in the reflector region and cooled by a dedicated cooling system. The "ex-pool" portion of the project, or Part 2 (which will be addressed in a subsequent license amendment), will involve the installation of (1) the hot cells necessary for the extraction of Mo-99 from the LEU target rods using the SGE process, and (2) the hot cells that would store the radioactive waste for decay and eventual shipment. It is also MURR's understanding, based on Currently, much of the supply ofMo-99 is produced using highly enriched uranium (HEU) targets. Because of global nonproliferation concerns, the future availability of HEU target material is uncertain. Research and development of suitable LEU targets for producing Mo-99 is the focus of various U.S. and foreign organizations.

conversations with the U.S. Nuclear Regulatory Commission (NRC), that the NRC will not approve the Part 1 amendment request for implementation until the NRC also approves the Part 2 amendment request. 2. Licensing Approach Background Representatives from MURR met with NRC staff on three separate occasions -April 27, 2015, June 2, 2016 and February 13, 2017 -to discuss MURR's approach to licensing the SGE experimental facility and provide an overview of the technical details of the project. At the end of the April 27, 2015 public meeting, the NRC suggested that MURR submit a written document outlining how MURR proposes to license the SGE experimental facility and the bases for that approach. By letter dated September 11, 2015, MURR provided a summary of MURR's licensing approach and the bases for requesting NRC approval pursuant to 10 C.F.R. § 50.90 by amending MURR's existing Facility Operating License. The September 11, 2015 letter outlined why MURR believed that the SGE activities at MURR would (1) be classified as an "experiment," (2) not require MURR to be categorized as a "testing facility," (3) not cause MURR to be categorized as a "production facility," (4) not require a construction permit per 10 C.F.R. § 50.90, and (5) not increase MURR's reactor core licensed power level limit of 10 MWt. During the June 2, 2016 public meeting, MURR presented a project update and restated our intended licensing approach, consistent with our September 11, 2015 letter. Also during that meeting, MURR proposed a 2-part licensing submission (In-Pool vs. Ex-Pool activities) for review efficiency such that the in-pool review could commence more promptly (while MURR completed its Part 2 information). MURR representatives left the June 2, 2016 meeting with the understanding that (1) the NRC was amenable to a 2-part submission based on the clear separation of in-pool and ex-pool activities, (2) the NRC would consider whether a construction permit was necessary based on any "alterations" or "material alterations" proposed by MURR, and (3) that the NRC staff felt strongly at the time that the ex-pool activities would require a "production facility" license. By letter dated January 17, 2017, MURR stated its position and associated bases for conclusions on the following four questions and requested formal responses from the NRC: 1. Will a Construction Permit be required? (10 C.F.R. § 50.23) 2. Will a change be required to MURR's licensed maximum operating power limit of 10 MWt due to the heat produced from the irradiation of the LEU targets? 3. Will a Production Facility license be required? 4. Will a change from a class 104c license (10 C.F.R. 50.21) to class 103 (10 C.F.R. 50.22) be required? (non-commercial vs. commercial) By letter dated February 2, 2017, the NRC stated that they had reviewed MURR's request, as well as the information provided by MURR and its partners, GA and Nordion (Canada), in the June 2, 2016 public meeting, and that the NRC needed additional information to better understand MURR's licensing approach. In this letter the NRC proposed discussion topics for a February 13, 2017 public meeting. On 2of10 February 13, 2017, MURR met with the NRC to further discuss the topics proposed by the NRC in the February 2, 2017 letter. In consideration of the February 13, 2017 meeting discussions, MURR provides herein its licensing approach for this amendment request [i.e., Part 1 ("in-pool")] and discusses the subsequent to-be submitted Part 2 (i.e., "ex-pool") license amendment request, which collectively address the SGE experiment at MURR. 3. Basis for the License Amendment -Part 1: "In-Pool" Activities 3.1 Construction Permit Not Required for "In-pool" Activities Implementation of Part 1 of the SGE experiment, or the "in-pool" activities, will be perfonned under an amendment to Renewed Facility Operating License No. R-103. MURR concludes that a construction permit is not required because implementation of the experiment will not necessitate a "material alteration" to the MURR facility per 10 C.F.R. § 50.92(a). The main components necessary for the "in-pool" activities will consist of LEU target assemblies, a separate and dedicated cooling system for heat removal (forced and natural convection), and instrumentation and control equipment to support these systems. NRC regulation 10 C.F.R. § 50.92(a), Issuance of Amendment states, in part: In determining whether an amendment to a license, construction permit, or early site permit will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses, construction permits, or early site permits to the extent applicable and appropriate. If the application involves the material alteration of a licensed facility, a construction permit will be issued before the issuance of the amendment to the license .... (Emphasis added.) MURR is not aware of any NRC definition of "material alteration." The NRC Staff considered defining "material alteration" in 1995 -1996, but apparently abandoned the effort because of "little regulatory need to clarify the term."2 As the NRC's Executive Director for Operations explained in SECY-96-024, In only one instance has a construction permit been issued before an amendment of an operating license, that is, an amendment to the operating license of a research reactor at the University of Maryland. The material alteration was the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement with those of a different design. The change rendered major portions of the safety analysis inapplicable.3 SECY-96-024, "Semiannual Status Report on Implementation of Regulatory Review Group Recommendations," at 5 (Feb. 2, 1996). Id. 3of10 The Staff concluded that there was an "apparent lack of need and industry or public interest in this topic." Material alterations over four decades ago at the University of Maryland which required a construction permit (issued March 25, 1970) prior to the issuance of the amendment (March 2, 1971) involved significant modifications to the reactor itself and its support systems. These modifications included the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement of those configurations with those of a different design, and major revisions to the associated accident analyses. Significant modifications also occurred at the Massachusetts Institute of Technology Reactor (MITR) where a construction permit was issued on April 9, 1973 prior to issuance of the license amendment (July 23, 1975). In this case the material alteration consisted of replacing the existing heavy-water moderated, cooled and reflected core with a light-water moderated, cooled, water reflected core, and modifications of installed systems to meet regulatory requirements. The change also rendered major portions of the safety analysis inapplicable. These examples involve significantly greater modifications than what will exist at MURR. In contrast to the University of Maryland and MITR examples, the SGE experiment does not involve the complete removal of major systems, structures, or components (SSCs) such as control rods or core instrumentation. The neutron producing core region of MURR will be minimally affected by the addition of the equipment for the experiment because the target rods will be located in the graphite reflector region of the reactor and used in a manner similar to other reflector region experiments. The installation and operation of this experiment utilizes a supporting target cooling system, and associated instruments and controls, that will better ensure that existing SSCs along with the design bases function of the SSCs as described in the MURR Safety Analysis Report (SAR) will not be adversely affected by the SGE experiment. As such, these changes do not constitute a significant modification to the facility when comparing them to previous NRC actions. 3.2 The Experiment Will Not Result In a Change to MURR's Operating Power Limit of 10 MW It is MURR's position that no change to the licensed steady-state operating power level limit of 10 MWt will be required. MURR uses the following calorimetric procedure to determine and maintain reactor power within the licensed limit using both primary and pool coolant system flow rates and differential temperatures: Reactor Power Level = [((total average primary coolant flow of each loop -primary demineralizer flow) x (Th -Tc)Primary) + ((average pool coolant flow) x (Th -Tc)r001)] x (a unit conversion factor) An additional backup calorimetric procedure, using secondary coolant system flow rate and differential temperature, is also performed: Reactor Power Level= [(secondary coolant flow) x (Th -Tc)secandary] x (a unit conversion factor) The SGE experiment will generate approximately 562 kWt of fission power from two (2) target assemblies. A dedicated cooling system will provide forced flow to each target assembly. This system 4of10 will draw coolant from the MURR reactor pool, pump it through dedicated heat exchangers which will lower the temperature of the coolant to the required value, then circulate the coolant through the target assemblies. The coolant leaving the target assemblies will then reenter the reactor pool at, or slightly above, the temperature of the bulk reactor pool water. Because the system is designed such that no heat is removed from the pool coolant system, the power generated by the target assemblies will not be part of MURR's 10 MWt reactor power level limit. MURR's secondary coolant system, which cools both the primary and pool coolant systems, will also provide the cooling water to the SGE experiment heat exchangers. The SGE experiment's cooling system is instrumented with flow, pressure and temperature monitoring equipment. Flow to each target assembly will be measured by flow elements. Temperature will be measured at four (4) separate locations: heat exchanger inlet, heat exchanger outlet, target assembly inlets and target assembly outlets. Secondary coolant to the SGE experimental facility cooling system will also be instrumented with heat exchanger inlet and outlet temperature elements and a flow element. MURR will use the same type of calorimetric procedure to determine SGE experimental facility power (individual target assembly power and total power) using coolant system flow rates and differential temperatures: Target Assembly A Power= [(target assembly coolant flow rate) x (target assembly differential temperature -Th -T charget Assembly A] x (a unit conversion factor) Target Assembly B Power= [(target assembly coolant flow rate) x (target assembly differential temperature -Th -Tcharget Assembly B] x (a unit conversion factor) SGE Experimental Facility Total Power = Target Assembly A Power + Target Assembly B Power An additional backup calorimetric procedure, using secondary coolant system flow rate and differential temperature across the SGE experiment heat exchanger, will also performed: SGE Experimental Facility Total Power = [(secondary coolant flow rate through SGE experimental heat exchanger) x (Th -Tc)secondary] x (a unit conversion factor) Redundant N-16 power level monitoring systems will also be installed on the primary coolant system piping exiting the reactor core. The N-16 monitoring systems will be calibrated at 100% reactor power operation without the SGE experimental facility operating. This will provide an additional method to the currently installed nuclear instrumentation to monitor reactor power. 4. Basis for the License Amendment-Part 2: "Ex-Pool" Activities This discussion is included with the Part 1 license amendment submittal to provide the NRC with a more complete picture of the total request (i.e., Part 1 and Part 2) at this time. In the Part 2 license amendment request, MURR plans to submit an application that includes a request for the Commission to approve an 5of10 amendment to its existing utilization facility license to expand the authorized activities to include possession, use, and operation of the facility under a combined production and utilization facility license. Presently, Renewed Facility Operating License No. R-103 authorizes MURR to "possess, use, and operate the facility as a utilization facility."4 The Commission has previously endorsed the use of a single, combined production and utilization facility license for medical isotope production facilities.5 The combined facility will be comprised almost entirely of SSCs that exist at MURR already. Only a few physical changes would be necessary to conduct the "ex-pool" activities -namely, installation of hot cells in the MURR facility. Hot cells routinely have been installed at MURR over the years without the need for a construction permit, and usually have been implemented via the 10 C.F.R. § 50.59 process.6 As explained further below, for the SGE experiment, such minor physical changes to the existing facility should not trigger the 10 C.F.R. § 50.lO(c) requirement for a construction permit to "begin" construction of the combined facility, nor should it trigger the 10 C.F.R. § 50.92(a) requirement for a construction permit for "material alterations" to the facility. 4.1 A Separate Production Facility License is Not Required The main components of the "ex-pool" phase of the SGE experiment will consist of hot cells and ancillary equipment to support those hot cells. LEU target rods irradiated during the "in-pool" phase will be transferred to the hot cells for processing and extraction of the Mo-99. NRC regulations at 10 C.F.R. § 50.1 O(b) require a license to "transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, or use any production or utilization facility." The NRC defines "production facility" in 10 C.F.R. § 50.2 to include "[a]ny facility designed or used for the processing of irradiated materials containing special nuclear material," subject to certain exceptions that are inapplicable to the SGE experiment. Notwithstanding these activities being part of an "experiment" as defined by the facility Technical Specifications, because of the quantity of batch material (and associated SNM) processed as part of the experiment in the hot cells, an NRC "production facility" license appears to be the most efficient pathway to licensing this SGE experiment. MURR is not aware of any NRC regulation or policy that would prohibit including both a production and utilization facility in one Part 50 license. In fact, as discussed below, the NRC previously considered such issues related to licensing medical isotope production facilities, and recommended using a single, combined production and utilization facility license in appropriate circumstances.7 4.1.1 The Commission Previously Has Approved the Use of a Combined License In 2007, Babcock and Wilcox (B&W) notified the Commission of its intent to seek a license to construct and operate a Medical Isotope Production System (MIPS). Prior to submitting its application, B& W Renewed Facility Operating License No. R-103 if 2.B.(2). See SRM-SECY-09-0101, Licensing ofa Babcock and Wilcox Medical Isotope Production System at page I (Oct. 9, 2009). See, e.g., Safety Evaluation Report, Renewal of the Facility Operating License for the University of Missouri-Columbia Research Reactor at pages 10-4 to 10-7 (Jan. 2017). SECY-09-0 I 0 I, Licensing of a Babcock and Wilcox Medical Isotope Production System at page 3 (July 9, 2009). 6of10 requested the NRC's views regarding several issues, including the possibility of licensing the MIPS under a single combined utilization and production facility license. In SECY-09-0101, the Staff provided the Commission with an assessment of licensing issues regarding the B& W request. In its evaluation of this request, NRC staff "examined whether it is legally feasible to issue one 10 C.F .R. Part 50 operating license for the entire MIPS system that incorporates both the utilization and extraction I purification portions of the facility."8 "The staff conclude[d] that there is no legal impediment under section 161.h of the AEA to issuing one 10 C.F .R. Part 50 operating license for entire MIPS facility (i.e., numerous reactors and one or more production facilities)."9 The NRC Staff also concluded that this licensing strategy should be considered on a case-by-case basis, and that it would need to be implemented "via Commission order." The Commission ultimately approved the Staff's approach.10 Similarly, no legal impediment exists to utilizing the current Part 50 license for utilization and production activities at MURR. 4.1.2 A Combined License is Appropriate for the SGE Experiment The SGE experiment is an appropriate scenario for using the single license approach. The benefit of this approach is that it provides a practical means to regulate common systems shared among and between the production and utilization facilities. The proposed "ex-pool" activities will share certain common systems with the existing reactor at MURR. Dividing up those systems among separate licenses would be unnecessarily burdensome with no commensurate enhancement to safety or security. Furthermore, approval of a single, combined production and utilization facility license would streamline the requirements applicable to MURR and promote regulatory efficiency. 4.2 A Construction Permit is Not Required for "Ex-Pool" Activities NRC regulations contain two provisions related to the potential requirement to obtain a construction permit. For new production or utilization facilities, 10 C.F.R. § 50.lO(c) requires a construction permit before breaking ground on a new facility (such as the MIPS, discussed above). More specifically, 10 C.F.R. § 50.lO(c) states: No person may begin the construction of a production or utilization facility on a site on which the facility is to be operated until that person has been issued . . . . a construction permit under this part .... For existing facilities, 10 C.F.R. § 50.92(a) requires issuance of a construction permit prior to approval of a license amendment if the modification would entail a "material alteration" of the facility. As explained below, the minimal changes proposed to MURR as part of the SGE experiment do not come remotely close to the "material alteration" threshold. Thus, the SGE experiment should not require a construction permit under either§ 50.lO(c) or§ 50.92(a). IO Id. Id. SRM-SECY-09-0101 at I. 7of10 4.2.1 10 C.F.R. § 50.lO(c) Does Not Apply in this Instance The NRC's definition of "construction" in§ 50.lO(a) (which applies to§ 50.lO(c)), is broad and ifread verbatim, could include nearly any physical alteration involving SSCs related to radiological health and safety. However, in practice, the NRC has not adopted this interpretation. Further, it appears that the § 50.10( c) requirement for a construction permit applies only to initial construction of a facility. By its own terms, the regulation prohibits "begin[ ning]" construction, but could be argued to be inapplicable after initial facility construction is complete. NRC-regulated power reactor licensees frequently engage in physical facility modifications that technically would fall within the definition of "construction," but based on research performed by MURR, traditionally have not been evaluated within the scope of § 50.lO(c) -we suggest, because the modification is not the "begin[ning]" of construction for the facility. Rather, modifications to existing facilities are covered under 10 C.F.R. § 50.92(a), in the license amendment context. Construction of MURR "was completed in substantial confonnity with the Construction Permit No. CPRR-68, issued on November 21, 1961."11 Essentially, the only physical change to the existing facility necessary to conduct the "ex-pool" activities is installation of the hot cells. The hot cells will share certain common systems with the existing reactor utilization facility, including exhaust ventilation, compressed air and electrical power. Accordingly, assuming the Commission approves an order authorizing a combined production and utilization facility license, part of the "production facility" apparatus already will exist. In other words, construction previously "began" under the 1961 Construction Permit, albeit via operation of the assumed Commission order approving the . existing apparatus for a dual purpose. Given the pre-existence of a portion of the production facility, it would be inconsistent to now state that a new construction would "begin" with the further modifications of the original structure, i.e., installation of the hot cells. Accordingly, MURR suggests that the language of§ 50.lO(c) does not apply to its license modification requests for inclusion ofMo-99 activities. The discussions herein are consistent with the NRC's statutory mandate for minimal regulation of research reactors. The Atomic Energy Act of 1954, as amended (AEA), accords special status to research facilities such as MURR. Specifically, Congress "directed" the Commission "to impose only such minimum amount of regulation" upon such facilities as is necessary to promote the common defense and security, protect public health and safety, and permit widespread and diverse research and development.12 Also note that the NRC has concluded in Part 52 space that its review of proposed facility alterations concurrent with the application is acceptable. In that context, the Commission has noted that, II 12 13 There is no safety or regulatory benefit in requiring the licensee to concurrently submit an application for a new Construction Permit in addition to a license amendment, inasmuch as NRC review of the alteration is assured.13 Renewed Facility Operating License No. R-103 'lJ J.B. Atomic Energy Act of 1954, as amended,§ 104.c (42 U.S.C. § 2134(c)). Licenses, Certifications, and Approvals for Nuclear Power Plants; Final Rule, 72 Fed. Reg. 49,352, 49,408 (Aug. 28, 2007). 8of10 Undoubtedly, NRC review of the proposed alteration for the SGE experiment "is assured" as part of the forthcoming "ex-pool" license amendment application. Accordingly, MURR concludes that a broad interpretation of § 50.lO(c) -one that would require the additional, cumulative step of obtaining a construction permit -would impose additional burdens of time and expense upon MURR (i.e., discourage research and development activities) without any safety or security benefit, contrary to the statutory commandment of AEA § 104.c. 4.2.2 The Proposed Hot Cell Installation is Not a "Material Alteration" The act of installing hot cells is not the type of"material alteration" contemplated in 10 C.F.R. § 50.92(a) for existing facilities. As explained in the "in-pool" discussion, above, MURR has identified only two circumstances where the NRC has required issuance of a construction permit for an amendment of an operating license for a non-power reactor facility. In one instance, the alteration included the complete removal of existing control rods, rod drive mechanisms, core instrumentation, and control room equipment and replacement with those of a different design; and in the other, replacing the existing heavy-water moderated, cooled and reflected core with a light-water moderated, cooled, heavy-water reflected core. Both modifications rendered major portions of the safety analyses inapplicable. Neither circumstance is remotely analogous to the mere installation of hot cells in the existing reactor building. Hot cells are routinely installed in MURR for various experiments. MURR has installed numerous hot cells throughout the life of the facility. Notably, the NRC has never required a construction permit for such installations. In fact, as part of the recent license renewal review, "the NRC staff observed the installation and use of additional hot cells and fume hoods," finding such installations were evaluated in accordance with 10 C.F.R. § 50.59 and adequately described in the MURR SAR.14 Even when Technical Specification changes are necessary for experiments conducted in hot cells, the NRC has not required a construction permit as a condition of the license amendment.15 Accordingly, the SGE experiment hot cell installation will not remotely affect a "material alteration" of the facility, as contemplated in 10 C.F.R. § 50.92(a), and a construction permit will not be required. 5. Summary In summary, MURR concludes that a combined production and utilization facility license is consistent with NRC policy. MURR also concludes that a construction permit is not required for the SGE experiment-related "ex-pool" activities under either 10 C.F.R. § 50.lO(c) or 10 C.F.R. § 50.92(a), because MURR is an existing facility, and consistent with previous NRC interpretations of its regulations, installation of hot cells would not be a "material alteration" of MURR, respectively. 14 15 See, e.g., Safety Evaluation Report, Renewal of the Facility Operating License for the University of Missouri-Columbia Research Reactor at 10-4 to 10-7 (Jan. 2017). See, e.g., Amendment No. 37 to Amended Facility License No. R-103 (Mar. 11, 2016) (ML16032A424). 9of10 If there are any questions regarding this response, please contact me at ( 573) 8 82-5118 or MeffertB@missouri.edu. I declare under penalty of perjury that the foregoing is true and correct. Sincerely, Bruce A. Meffert Reactor Manager xc: Reactor Advisory Committee Reactor Safety Subcommittee Isotope Use Subcommittee Dr. Hank Foley, Interim Chancellor ENDORSEMENT: Reviewed and Approved Ralph A. Butler, P.E. Director Dr. Mark Mcintosh, Vice Chancellor for Research, Graduate Studies and Economic Development Mr. Alexander Adams, U.S. Nuclear Regulatory Commission Mr. Geoffrey Wertz, U.S. Nuclear Regulatory Commission Mr. Johnny Eads, U.S. Nuclear Regulatory Commission Attachments: ,-1. Attachment 1 -License Amendment Request to Implement Selective Gas Extraction (SGE) Target Experimental Facility (TEF) at the University of Missouri Research Reactor 2. Attachment 2 -Revised and New Technical Specifications 3. Attachment 3 -Codes and Standards 4. Attachment 4-GA Design Report No. 30441R00021, "Target Assembly Thermal Analysis" 5. Attachment 5 -GA Design Report No. 30441R00022, "Source Term Analysis Design Calculation Report" 6. Attachment 6 -GA Design Report No. 3 0441 R00031, "Mo-99 Target Assembly Nuclear Design for Once-Through Operation" 7. Attachment 7 -GA Design Report No. 30441R00032, Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report" 8. Attachment 8-GA Design Report No. 30441R00017, "ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report" 10of10 JACQUELINE L. MATYAS My Commission Expires Man:h 26, 2019 Howard County Commission #1563430&

+GENERAL ATOMICS AND AP/t=ll.IATED COMPANIES GENERAL ATOMICS AFFIDAVIT OF KEITH E. ASMUSSEN I, Keith E. Asmussen, Director, Licensing, Safety and Nuclear Compliance, General Atomics, do hereby affirm and state: (1) I have been delegated the function of reviewing information described in paragraph 3 which General Atomics requests be withheld from public disclosure or publication and I am authorized to execute this affidavit on behalf of General Atomics. (2) The affidavit is submitted under the provisions of 1 OCFR Part 2.390 in order to withhold documents containing confidential commercial and proprietary information (as set forth in paragraph 3 following) of General Atomics from public disclosure or publication. (3) General Atomics (GA) has partnered with The University of Missouri Research Reactor (MURR) and Nordion to develop the Reactor-Based Mo-99 Supply System (RB-MSS) Project using its Selective Gas Extraction process. The information sought to be withheld is related to the analysis, design, development and licensing of General Atomics' technology for deployment at MURR. This information is contained in the following documents submitted as supporting information for MURR's "License Amendment Request to Implement Selective Gas Extraction Target Experimental Facility at the University of Missouri Research Reactor." License Amendment Request TITLE Attachment No. License Amendment Request to Implement Selective Gas 1 Extraction (SGE) Target Experiment Facility (TEF) at the University of Missouri Research Reactor 4 GA Report 30441 R00021: "Target Assembly Thermal Analysis" 5 GA Report 30441R00022: "Source Term Analysis Design Calculation Report" 6 GA Report 30441 R00031: "Mo-99 Target Assembly Nuclear Design for Once-Through Operation" GA Report 30441 R00032: "RELAP Accident Analysis and 7 FRAPTRAN Target Rod Transient Analysis Design Calculation Report" 8 GA Report 30441 R00017: "ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report" 3550 GENERAL ATOMICS COURT.SAN DIEGO, CA 92121-1122 PO BOX 85608, SAN DIEGO, CA 92186-5608 (858) 455-3000 Essentially each and every page of these documents contains proprietary material developed by General Atomics, and for General Atomics by its partners MURR and Nordion that is "confidential", "proprietary", "business sensitive" and/or "trade secret". Nonproprietary redacted versions of these documents are being simultaneously provided where the proprietary information on each page has been appropriately redacted. (4) In making this application for withholding of proprietary information of which it is the owner, General Atomics relies upon the exemption from disclosure set forth is the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential." The material for which exemption from disclosure is hereby sought is all "confidential commercial information," and/or also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704F2d1280 (DC Cir. 1983). (5) Some examples of categories of information which fit into the definition of proprietary information are: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Atomics' competitors without license from General Atomics constitutes a competitive economic advantage over other companies; b. Information which, if used by a competitor, would reduce his or her expenditure of resources or improve his or her competitive position in the design, manufacture, shipment, installation, assurance or quality, or licensing of a similar product. c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection. (6) The information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence, is of a sort customarily held in confidence by General Atomics, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by General Atomics. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. * (7) Initial approval of proprietary treatment of a document is made by the manager of the originating business unit, the person most likely to be acquainted with the value and sensitivity of the information in* relation to industry knowledge. Access to such documents within General Atomics is limited on a "need to know" basis. Disclosures outside General Atomics are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary non-disclosure agreements for protecting the information from further disclosure. (8) The information classified as proprietary was developed and compiled by General Atomics at a significant cost to General Atomics. This information is classified as proprietary because it contains detailed data and analytical results not available elsewhere. This information would provide other parties, including competitors, with information from General Atomics technical database and the results of evaluations performed by General Atomics. Release of this information would improve a competitor's position without the competitor having to expend similar resources for the development of the database. A significant effort has been expended by General Atomics to develop this information. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to General Atomics' competitive position. The information is part of General Atomics' selective gas extraction technology base, and its commercial value extends beyond the original development cost. The research, development, engineering, and analytical costs associated with General Atomics' unique selective gas extraction sy.stem and process comprise a substantial investment of time and resources by General Atomics. GENERAL ATOMICS Keith E. Asmussen Director, Licensing, Safety and Nuclear Compliance A notary public or other officer completing this certificate verifies only the identity of the individual who signed the document to which this certificate is attached, and not the truthfulness, accuracy, or validity of that document. State of California County of San Diego On 3 / J5 { l 7 before me, Joyce E. Zachman, Notary Public, personally appeared Keith E. Asmussen, who proved to me on the basis of satisfactory evidence to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his aLJthorized capacity, and by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument. I certify under PENAL TY OF PERJURY under the laws of the State of California that the foregoing paragraph is true and correct. WITNESS my hand and official seal. i98tlf re of Not) JOYCE E. ZAQHMAN Cominihion #"2021697 * . Notary Pubilc .*. l z , *

San D'i,ego County -) , ..

Ml c:omm. Expires Apr*2s; 2011*1 ******************

ATTACHMENT 1 LICENSE AMENDMENT REQUEST TO IMPLEMENT SELECTIVE GAS EXTRACTION (SGE) TARGET EXPERIMENTAL FACILITY (TEF) AT THE UNIVERSITY OF MISSOURI RESEARCH REACTOR ATTACHMENT 1 TABLE OF CONTENTS TABLE OF CONTENTS .............................................................................................................................. 2 LIST OF FIGURES ...................................................................................................................................... 3 LIST OF TABLES ........................................................................................................................................ 7 1. Selective Gas Extraction Target Experimental Facility Overview .................................................... 10 1.1 Proposed Experiment Description ......................................................................................................... 10 1.2 Normal Operation .................................................................................................................................. 12 1.3 Design for Loss of Forced Cooling ........................................................................................................ 13 1.4 Target Experimental Facility Safety Design Approach ......................................................................... 13 1.5 Design Bases .......................................................................................................................................... 15 1.6 Target Assembly Maximum Power ....................................................................................................... 16 1. 7 Target Rod Peak Linear Power .............................................................................................................. 16 1.8 Codes and Standards for Design, Fabrication and Operations ............................................................... 17 2. Detailed Selective Gas Extraction Target Assembly Description and Structural Analysis .............. 17 2.1 TargetAsseinbly .................................................................................................................................... 17 2.2 Structural Analysis of Target Assembly ................................................................................................ 28 3. Target Cooling System ...................................................................................................................... 36 3.1 Functions ......................... : ..................................................................................................................... 36 3.2 Target Cooling System Principal Design Parameters ............................................................................ 37 3.3 Target Cooling System Design .............................................................................................................. 38 3 .4 Target Cooling System Performance .................................................................................................... .45 4. Instrumentation and Control System ................................................................................................. 52 4.1 Summary Description ............................................................................................................................ 52 4.2 Target Cooling System Control System Description ............................................................................. 54 4.3 Target Cooling System Protection System Description ......................................................................... 55 5. Target Assembly Nuclear Design Analysis .......... ; ........................................................................... 65 5.1 Analytical Methods ................................................................................................................................ 65 5.2 Target Assembly Physics Model ........................................................................................................... 65 5.3 Target Assembly Flux and Power .......................................................................................................... 76 6. Target Assembly Thermal Hydraulic Design Analysis ................................................................... 103 6.1 Thermal-Hydraulic Design Basis ......................................................................................................... 104 Page 2of190 ATTACHMENT 1 7. In-Pool Target Transfer Syste1n ...................................................................................................... 124 7 .1 Cartridge Loading/Unloading Station Design ...................................................................................... 124 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing ............................... 127 8. Radiological Protection Evaluation for the SGE Target Experimental Facility Operations ........... 128 8.1 Airborne Sources ................................................................................................................................. 129 8.2 Liquid Sources ..................................................................................................................................... 129 8.3 Solid Sources ....................................................................................................................................... 131 8.4 Radioactive Waste Management Program ........................................................................................... 132 9. Conduct of Operations .................................................................................................................... 134 9.1 Procedures ............................................................................................................................................ 134 10. Target Experimental Facility Accident Analyses ............................................................................ 139 10.1 Target Experimental Facility Maximum Hypothetical Accident ......................................................... 139 10.2 Insertion of Excess Reactivity ............................................................................................................. 155 10.3 Control Blade Withdrawal ................................................................................................................... 160 10.4 Loss of Target Coolant ........................................................................................................................ 163 10.5 Pipe Break Locations Out of the Reactor Pool .................................................................................... 164 10.6 Pipe Break Locations in the Reactor Pool ........................................................................................... 170 10.7 Loss of Target Flow ...................................................... : ...................................................................... 175 10.8 Mishandling of Target Cartridge or Target Rods ................................................................................. 177 10.9 Loss of Primary Coolant Flow ............................................................................................................. 177 10.10 Loss of Primary Coolant ...................................................................................................................... 178 10.11 Loss of Pool Coolant.. .......................................................................................................................... 178 10.12 Loss ofOffsite Electrical Power .......................................................................................................... 178 11. Technical Specifications ................................................................................................................. 178 12. Proposed Confirmatory Testing ..................... : ................................................................................ 178 12.1 Summary Description of Planned Tests ............................................................................................... 179 13. References ....................................................................................................................................... 188 LIST OF FIGURES Figure 1 Selective Gas Extraction Process Scope, Functional Relationships and Interfaces ....................... 11 Figure 2 Layout of SGE Experimental Facility in the MURR Graphite Reflector Region and Containment. ............................................................................................................................... 11 Page 3of190 ATTACHMENT 1 Figure 3 Location of Target Assemblies in MURR Graphite Reflector Positions No. 5A and No. 5B ....... 12 Figure 4 Target Asse1nbly ....................................................................................... 18 Figure 5 Illustration of Target Housing Elevation and Plan Views ............................................................. 19 Figure 6 Cartridge Configuration and Target System Section View ........................................................... 20 Figure 7 Cartridge Upper and Lower Sections ............. ............................................................................... 21 Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum ........ 22 Figure 9 Target Rod Arrangement ............................................................................................................... 23 Figure 10 Neutron Absorber Section View .................................................................................................. 24 Figure 11 Inlet and Outlet Plenums with Method of Attachment (water flow is shown in blue) ................ 25 Figure 12 Upper Target Cartridge Arrangement (water flow is shown in blue) .......................................... 26 Figure 13 Target Pellet Geometry ................................................................................................................ 27 Figure 14 Allowable Strength vs Temperature for Al 6061T6 and SS316L ................................................ 29 Figure 15 Pellet and Cladding Details .......................................................................................................... 30 Figure 16 Allowable Stress .................... 32 Figure 17 Fatigue Chart ......................................... 32 Figure 18 Deflection Due to Target Rod Bowing for Worst-Case Front-to-Back Power Skew .................. 35 Figure 19 Process Flow Diagram of Target Cooling System ....................................................................... 39 Figure 20 Elevation View of Target Cooling System .................................................................................. 40 Figure 21 Target Cooling System Superimposed onto MURR Reactor Pool and Biological Shield ........... 41 Figure 22 Schematic for the Water Cooling Module ................................................................................... 42 Figure 23 Target Cooling System ................................................................................................................ 43 Figure 24 State Points within the Target Cooling System ........................................................................... 45 Figure 25 Target Cooling System P&ID ........ , ............................................................................................. 46 Figure 26 Location and Type of Supports on Above Pool Piping ............................................................... 47 Figure 27 Location and Type of Support on In-Pool Piping ........................................................................ 48 Figure 28 Secondary Cooling for Target Cooling System ........................................................................... 50 Figure 29 Electric Power Supply to TCS Pumps and I&C UPS .................................................................. 52 Figure 30 Target Cooling System Control System Architecture .................................................................. 54 Figure 31 Target Cooling System Control Panel ......................................................................................... 55 Figure 32 Target Cooling System Protection System Relay Inputs to the MURR Reactor Safety System ........................................................................................................................................ 60 Figure 33 TDHRVs -TCS Pump Interlock Circuit.. ................................................................................... 62 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane ................................................................................................................................... 66 Page 4of190 ATTACHMENT 1 Figure 35 MCNP6 Model of Target Assembly at Axial Mid-Plane ............................................................ 66 Figure 36 Target Rod Numbering for. Target Assemblies (baseline model) ........................................ 67 Figure 37 Model of the Target Cartridge .......................................................... 67 Figure 38 MURR Control Blade Travel between January 13, 2014 and September 15, 2015 .................... 69 Figure 39 Target Rod I Axial Neutron Flux Distribution ............................................. 77 Figure 40 Target Rod. Axial Neutron Flux Distribution ........................................... 77 Figure 41 Target Assembly Azimuthal Neutron Flux Distribution .............................................................. 78 Figure 42 Power Envelope of the Base Target Loading for the Extreme Bumup Core Case ...................... 79 Figure 43 Power Envelope of the Base Target Loading for the Maximum Bum up Core Case ................... 79 Figure 44 Target Assembly Pellet Linear Power Distribution .................... 80 Figure 45 Variation of Target Power vs. Control Blade Position ................................................................ 83 Figure 46 Variation of Peak Linear Power vs. Control Blade Position ........................................................ 83 Figure 47 Variation of Driver Fuel Element Peaking Factor vs. Control Blade Position ............................ 84 Figure 48 Axial Neutron Flux ....................................................................................................................................... 89 Figure 49 -Axial Neutron Flux ---=********************************************************************************************************************************90 Figure 50 Target Assembly Azimuthal Neutron Flux ......................................................................................................................... 90 Figure 51 Axial Neutron Flux ....................................................................................................................................... 91 Figure 52 Axial Neutron Flux for ....................................................................................................................................... 91 Figure 53 Target Assembly Azimuthal Neutron Flux ......................................................................................................................... 92 Figure 54 Figure 55 ..................................................................................................... 94 Figure 56 Target Assembly Pellet Linear Power Distribution ..................................................................................................... 95 Figure 57 Power Envelope of the Staggered Loading ...................................................................................................... 96 Figure 58 Power Envelope of the Staggered Loading ...................................................................................................... 96 Figure 59 Target Assembly Pellet Linear Power Distribution ...................................................................................................... 97 Figure 60 -Axial Neutron Flux for Partial Loading Case ........................................................ 98 Page 5of190 ATTACHMENT 1 Figure 61 Axial Neutron Flux for Partial Loading Case ...................................................... 99 Figure 62 Target Assembly Azimuthal Neutron Flux for Partial Loading Case .......................................... 99 Figure 63 Power Envelope for Partial Loading for the Extreme Bumup Core Case ................................. 100 Figure 64 Power Envelope of the Partial Loading for the Maximum Bumup Core Case .......................... 100 Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case ........................... 101 Figure 66 Change oflrarget Assembly Power .................................................... 102 Figure 67 Vapor Fraction at Cladding Wall for Worst-Case Conditions in FLUENT RPI Wall Boiling Model ........................................................................................................................... 107 Figure 68 Heat Flux and CHFR as a Function of Axial Location for Peak Power Target Rods ................ 108 Figure 69

Thermal Conductivity and Thennal Expansion Coefficient --**************************************************************************************************************************************111 Figure 70 -Thennal Conductivity and Radial Thermal Expansion Coefficient.. ....................... 111 Figure 71 -Thennal Conductivity in Small Gaps ........................................................................... 112 Figure 72 Target Rod Loading/Unloading/Storage Location ..................................................................... 124 Figure 73 Representation of the Cartridge Loading/Unloading Station ..................................................... 125 Figure 74 Target Rod Remote Handling Tool.. .......................................................................................... 127 Figure 75 Reactor Power, Fuel and Cladding Temperatures vs. Time for a Positive Reactivity Step Insertion of 0. 006 Lik/k ............................................................................................................. 15 6 Figure 76 Cladding Strains at Peak Pellet Location During a Positive 0.006 Lik/k Reactivity Insertion ... 158 Figure 77 Peak Target Pellet Temperature During a Positive 0.006 Lik/k Reactivity Insertion ................. 158 Figure 78 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a Positive 0.006 Lik/k Reactivity Insertion .................................................................................. 159 Figure 79 Power Transient During a 0.0003 .ilk/k per Second Reactivity Insertion .................................. 160 Figure 80 Cladding Strains at Peak Pellet Location During a 0.0003 .ilk/k per Second Reactivity Insertion .................................................................................................................................... 161 Figure 81 Peak Target Pellet Temperature During a 0.0003 .ilk/k per Second Reactivity Insertion .......... 162 Figure 82 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a 0.0003 Lik/k per Second Reactivity Insertion ........................................................................... 163 Figure 83 Pipe Break Locations Out of the Reactor Pool .......................................................................... 164 Figure 84 Mass Flow Transient During a LOCA in Air Without Decay Heat Removal Valves Opening .................................................................................................................................... 166 Figure 85 Target Power During a LOCA in Air Without Decay Heat Removal Valves Opening ............. 167 Figure 86 Coolant Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening .................................................................................................................................... 168 Figure 87 Maximum Cladding ID Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening ......................................................................................................... 168 Figure 88 Mass Flow Transient During a LOCA in Air With Decay Heat Removal Valves Opening ...... 169 Page 6of190 ATTACHMENT 1 Figure 89 Maximum Cladding ID Temperatures During a LOCA in Air With Decay Heat Removal Valves Opening ........................................................................................................................ 170 Figure 90 Pipe Break Location in the Reactor Pool.. ................................................................................. 171 Figure 91 Mass Flow Transient During a LOCA in the Reactor Pool ....................................................... 172 Figure 92 Target Power During a LOCA in the Reactor Pool ................................................................... 173 Figure 93 Maximum Cladding ID and Coolant Temperatures During a LOCA in the Reactor Pool ........ 174 Figure 94 Cladding OD Temperature Profile in Target Rod. During a LOCA in the Reactor Pool ...... 174 Figure 95 Cladding Fractional Strains in Rod. during a LOCA in the Reactor Pool ............................. 175 Figure 96 Mass Flow Transient During Loss of Pump Flow ..................................................................... 176 Figure 97 Maximum Cladding ID and Coolant Temperatures During Loss of Pump Flow ...................... 177 Figure 98 Conceptual Design of Test Capsule ........................................................................................... 180 Figure 99 Visualization of Potential Pellet Deformation over Course oflrradiation (not to scale) ........... 181 Figure 100 Conceptual Schematic of CHF Test Flow Section ................................................................... 184 Figure 101 Uninstrumented Critical Heat Flux Test Section in Low Pressure Flow Loop Rig .................. 185 Figure 102 GA Test Setup Conceptual Arrangement ................................................................................. 186 LIST OF TABLES Table 1 Temperature Limits for Target Assembly Design ........................................................................... 14 Table 2 Design Calculation Reports ............................................................................................................. 15 Table 3 Target and Filler Rod Dimensions (cold) ........................................................................................ 23 Table 4 FRAP CON Results Su1nmary ......................................................................................................... 31 Table 5 Maximum Irradiation Damage to Target Cartridge Materials ........................................................ 33 Table 6 Target Cooling System Design Parameters for Two Targets ......................................................... 38 Table 7 Major Component Specifications .................................................................................................... 44 Table 8 Results of Rigid Structure Seismic Analysis ................................................................................... 49 Table 9 Definition of MURR Driver Core Bumup States (MWD) .............................................................. 68 Table 10 Critical Control Blade Positions .................................................................................................... 70 Table 11 Expected Control Blade Travelling Range for Target-Loaded Core ............................................. 70 Table 12 Keff and Reactivity Insertion Values for Single Target Assembly ................................................ 71 Table 13 MURR Core Maximum Power Peaking Due to Target Assembly Loading ................................. 72 Table 14 Reactivity Coefficients of the Reactor Core with. Fresh Target Assemblies ........................ 73 Table 15 Reactivity Control Device Worth with. Fresh Target Assemblies ......................................... 74 Table 16 Kinetic Parameters of the Reactor Core with and without Target Assemblies ............................. 75 Page 7of190 ATTACHMENT 1 Table 17 Component Radiation Heating Due to Target Assembly Loading for Maximum Burnup Core ............................................................. '. ............................................................*................. 76 Table 18 Calculated Target Power Level and Linear Power ........................................................................ 78 Table 19 Target Assembly Power and Core Peaking Factors ...................................................................... 82 Table 20 Effect of Regulating Blade on Pellet Peak Linear Power and Target Assembly Power ............... 85 I Table 21 Material Impurities for Uncertainty Analysis ............................................................................... 85 Table 22 Uncertainties in Core Eigenvalue (Bias) ....................................................................................... 87 Table 23 Uncertainties in Peak Linear Power (Bias) ................................................................................... 88 Table 24 Uncertainties in Target Power (Bias) ............................................................................................ 88 Table 25 Calculated Target Power Level and Pellet Linear Power ................... 93 Table 26 Calculated Target Power Level and Pellet Linear Power .............................................................. 98 Table 27 Material Content and Burnup of the Base Target Loading for Average Burnup Core State ...... 101 Table 28 Variation of Coolant Inlet Temperature with Active Target Rod Count... .................................. 110 Table 29 Predicted Thermal Perfonnance for -Target Rods, 10 MWt Reactor Power, .. *.......................................................................................................................................... 113_ Table 30 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, * *.......................................................................................................................................... 115 Table 31 Predicted Thennal Performance for -Target Rods, 10 MWt Reactor Power, .. *.......................................................................................................................................... 117 Table 32 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, * *.......................................................................................................................................... 119 Table 33 Predicted Thennal Performance for -Target Rods, Worst-Case Operations .................. 121 Table 34 Predominant Radionuclides in the MURR Pool Coolant and Their Measured Concentrations at 10 MW .................................................... ; .................................................... 130 Table 35 Representative Radioactive Sources at MURR ........................................................................... 132 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI ................................................................................................................................... 134 Table 37 Standard Operating Procedures ................................................................................................... 135 Table 38 Activity of Volatile Fission Products in the Gap Gas of the Hottest Target Rod ....................... 141 Table 39 Iodine and Noble Gas Activities Released to MURR Reactor Pool ........................................... 142 Table 40 Iodine Concentrations in Pool Water .......................................................................................... 142 Table 41 Average Iodine Concentrations in the Containment Building Air During .................................. 143 Table 42 Average Noble Gas Concentrations in the Containment Building Air during the 5 Minute Evacuation Period ..................................................................................................................... 144 Table 43 Derived Air Concentration Values and 5-Minute Exposures for Iodine ..................................... 146 Table 44 Derived Air Concentration Values and 5 Minute Exposures -Noble Gases .............................. 147 Page 8of190 ATTACHMENT 1 Table 45 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building ................... 147 Table 46 Average Containment Building Leakage Rate ............................................................................ 149 Table 4 7 Average Iodine Concentrations in Air Exiting the Exhaust Stack .............................................. 150 Table 48 Noble Gas Concentrations in Air Exiting the Exhaust Stack ...................................................... 151 Table 49 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Iodine .................................................................. 153 Table 50 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases ........................................................ 153 Table 51 Dose from Iodines and Noble Gases in the Unrestricted Area .................................................... 154 Table 52 Radiation Shine through the Containment Building ................................................................... 154 Table 53 Capsule Dimensions and Attributes ............................................................................................ 179 Table 54 Irradiation Test Schedule ............................................................................................................ 181 Table 55 Critical Heat Flux Test Schedule ................................................................................................ 184 Table 56 Testing Specifications ................................................................................................................. 187 Table 57 System Integration Test Schedule ............................................................................................... 187 Page 9of190 ATTACHMENT 1 1. Selective Gas Extraction Target Experimental Facility Overview 1.1 Proposed Experiment Description The Selective Gas Extraction (SGE) experiment employs a first-of-a-kind concept for radioisotope production. It is a reactor-driven, low-enriched uranium (LEU)-based system that selectively removes specific isotopes of interest, viz., molybdenum-99 (Mo-99), in gaseous fonn, that are produced from fission during irradiation of Zircaloy-4 clad target rods containing LEU. The LEU is in the form of uranium dioxide (U02) pellets that are nominally enriched to -in the isotope uranium-235 (U-235). The target rods are contained in -cartridges that ensure uniform cooling water flow around the target rods and will be located in the graphite reflector region of the University of Missouri Research Reactor (MURR). The SGE Target Experimental Facility (TEF) will be operated by MURR staff in concert with MURR's routine reactor operations. During SGE experiment operation, one (1) or two (2) target cartridges, holding. target rods each, are placed in permanently installed support assemblies in the reactor graphite reflector region. Fission product isotopes, including Mo-99, are generated during target irradiation. At the end of irradiation. -'and following a short period of cooling (to reduce decay heat), the target rods are removed from their respective cartridge and transferred to a loading/unloading and storage location within the reactor pool. Subsequent target rod transfer to a bank of hot cells inside the reactor containment building will be addressed in the Part 2 license amendment submission. Figure 1 shows a functional block diagram of the SGE experiment overview for this Part 1 License Amendment submission as well as for the Part 2 submission. The in-pool (IP) portion of SGE process, the TEF, is the focus of this reactor License Amendment request. A description of these IP systems and associated analyses are presented in this document as follows: 1. Target System including the Target Rods, Cartridge, and Housing -) 2. Target Cooling System (TCS) 3. Cartridge Loading/Unloading and Storage Station 4. Instrumentation and Control (I&C) Systems 5. Analyses of Potential Accidents and their Impact on the MURR Facility The remaining SGE process (ex-pool) systems and associated safety analyses will be described in the Part 2 submission. Figure 2 and Figure 3 show the layout of the TEF irradiation systems within the MURR reactor pool and containment building. Page 10of190 ATTACHMENT 1 Figure 1 Selective Gas Extraction Process Scope, Functional Relationships and Interfaces CARTRIDGE LOADL'"G/li:\"LOADL'"G STATIO:S CARTRIDGE WITH TARGET RODS TARGET Figure 2 Layout of SGE Experimental Facility in the MURR Graphite Reflector Region and Containment Page I I of 190 ATTACHMENT 1 Figure 3 Location of Target Assemblies in MURR Graphite Reflector Positions No. SA and No. SB 1.2 Normal Operation During nonnal operation, cartridge(s) are loaded in the loading station shown in Figure 2. The cartridge(s) is then inserted into the target housing(s) located in graphite reflector positions while the reactor is shut down. When the reactor is restarted, neutrons from the reactor will fission the LEU in the target rods generating fission are cooled by forced flow from the Target Cooling System (TCS) both when the reactor is operating and to remove decay heat during routine reactor shutdowns. The heat removed from the T As is rejected via the TCS to the existing facility secondary coolant system without significantly affecting the reactor bulk pool temperature. The nominal operation cycle will load and remove .arget rods per week for processing. At the end of the irradiation cycle, the reactor is shut down and the TA is allowed to cool by forced flow from the TCS . At this time the level of stored and decay heat in the cartridge is sufficiently low that direct conduction to the reactor pool water is sufficient to maintain the target rods in a safe condition. The cartridge(s) is then moved to the loading/unloading station and the target rods are transferred to in-pool storage or a shielded transfer cask which has been moved to an underwater position near the loading/unloading station for safe transfer. Meanwhile a previously prepared fresh cartridge(s) is loaded into the TA positions Page 12 of 190 ATTACHMENT 1 1.3 Design for Loss of Forced Cooling Loss of forced cooling (LOPC) to the TA could occur by a power failure, pump failure or coolant pipe break. If a LOPC event occurs, a -signal ( -is sent from the SGE TEP to the MURR reactor safety system to scram the reactor. Decay heat removal from the target rods is provided by pool water 1.4 Target -Experimental Facility Safety Design Approach The safety objectives for the SGE TEP are (1) minimization of exposure of occupational workers to radiation from nonnal system operation, (2) prevention of release of radionuclides to the MURR reactor containment building and the environment from operation of the system, and (3) control of target rod cladding temperatures to prevent cladding failure leading to a fission product release to the reactor containment building and potentially, to the environment. To meet these objectives, the SGE TEP has been designed to the following top-level requirements to ensure that the safety objectives are met:

The target rods and cartridges shall be capable of sustained operation with the reactor operating at a steady-state power level of 10.0 MW1 and through anticipated operating transients where the reactor power reaches its scram power limit, without jeopardizing the integrity of barriers that would release radionuclides to the reactor pool water, target cooling water, or the reactor containment building.

The target cartridges together shall not increase the reactivity worth by more than 0.006 Lik/k when inserted into positions of the graphite reflection region.

The target rods shall be hennetically sealed in cladding that will serve as a barrier against the release of radioisotopes to the reactor pool water, target cooling water, or the reactor containment building.

In case of a loss of forced flow, pool water through the TA shall provide an adequate backup means of removing target decay heat without damage to the cladding.

Normal cooling shall ensure that the minimum critical heat flux ratio (MCHPR) using the Bernath correlation shall not fall below 2.0 for nonnal operation.

The SGE TEP shall interface with the MURR reactor safety system to initiate a reactor scram under any condition that threatens the barriers against the release of radionuclides to the reactor pool water, target cooling water, or the reactor containment building.

The SGE TEP Instrumentation and Control (I&C) System shall incorporate instrumentation to ensure that essential system variables are monitored and are within specified operating ranges at all times. Page 13 of 190 ATTACHMENT 1

The SGE TEF I&C System shall be equipped with a reliable and redundant system that monitors the operation of systems required for safety. Upon detection of conditions that indicate an anomalous state, failure or impending failure of a barrier against the release of radionuclides to the reactor pool or the containment building, the system shall cause a reactor scram, or generate appropriate alanns to cause operator action to the anomalous condition.

The SGE TEF shall satisfy the facility limits for any releases of radioactivity to the environment. Target and cladding design temperature limits are shown in Table 1. Table 1 Temperature Limits for Target Assembly Design Component Normal Operation Transients target rod cladding2 The following design features mm1m1ze the likelihood of the target cladding exceeding the design temperature limits during irradiation and processing, and causing a release of radioactivity to the environment:

The U-235 loading in the target rod is limited to ensure the T As remain substantially subcritical under normal operation and all credible off-nonnal conditions.

The MURR reactor will scram on low or high TA cooling water flow or loss of heat sink (secondary coolant flow) thereby limiting heat generation in the target.

The TA is designed to permit cooling by natural convection using reactor pool water, in the event of a LOFC transient.

The heat flux from the target rods is limited by design to ensure the temperature of the -cladding does not exceed the design limits.

The target rod cladding serves as a barrier to the release of radionuclides. A breach of this barrier will be detected by the reactor pool coolant system radioactivity monitoring system. Page 14of190 ATTACHMENT 1

In the unlikely event of a leak of radionuclides from the target rods, the reactor pool serves to reduce the amount of both Iodine and noble gas release due to plating out of Iodine, decaying of noble gases by slowing down the time it takes the gas to rise to the pool surface, and gas solubility. 1.5 Design Bases During the course of target operation the total thermal power and power distribution will depend upon the combination of several independent variables including bumup and location of the fuel elements, age and position of the control blades, age of the beryllium reflector, time in reactor operating cycle and time in target operating cycle. This leads to a near-continuum of core and target states. The target nuclear design analyses are performed for a selected set of conditions that typify the variation in target operating states. No one set of conditions can fonn a conservative basis for all the safety analyses, but rather the safety design basis was created from a composite of several states to give the conservative worst-case conditions for evaluating the safety performance of the TEP operations. Design bases, including all assumptions for the design bases, as well as documentation of software verification calculations that form the bases of the design and safety information presented in the License Amendment are provided in the reports listed in Table 2 below. Table 2 Design Calculation Reports REPORT NO. TITLE 30441R00017 ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report 30441R00019 Target System Cooling Calculation Report 30441R00021 Target Assembly Thermal Analysis 30441R00022 Source Term Analysis Design Calculation Report 30441R00030 Mo-99 Target Cooling System Seismic Analysis Design Calculation Report 30441R00031 Mo-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00032 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441R00033 Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations 30441R00035 Cooling of the MURR Beryllium Reflector 30441R00038 Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report 30441M00043 ANSYS Thermal Model of RB-MSS Target Rod Page 15 of 190 ATTACHMENT 1 1.6 Target Assembly Maximum Power For nonnal irradiation operation, the maximum target power detennines the coolant flow rate necessary to meet the required TA outlet temperature. The maximum target power also establishes the safety design bases for the fission product source tenns and for removal of the residual decay heat from the target rods in case normal flow is lost. The maximum target power therefore establishes the decay heat loads and source tenns for safety evaluations. This value corresponds to MURR core power at 10 MW1 and a maximum core bum up case -* This case produces a peak power in fuel element -Nevertheless, this value is conservatively used for the afterheat level and fission product source terms The worst-case control blade tilting and age is assumed (0 years for blades 'A' and 'D' and 8 years for blades 'B' and 'C') and the beryllium reflector age is assumed to be 0 years. This worst-case value conservatively corresponds to a control blade withdrawal position at The estimated 2cr statistical uncertainty in the total power for two (2) T As is -* The total uncertainties of the key performance parameters were estimated as a product of statistical error and mean-square (RMS) of uncertainties due to fabrication density, enrichment, and target rod position. Uncertainty in impurities was conservatively not included because it is always negative. An additional heat load for the required cooling is the structural heating from the cartridge and neutron absorber shield. The heat from the cartridge and neutron absorber shield is The maximum fission power used for the safety analyses is 1. 7 Target Rod Peak Linear Power For normal operation, the peak target rod linear power determines the combination of coolant flow velocity, temperature and pressure required to achieve adequate heat transfer margin to prevent film boiling and centerline pellet melting. For abnonnal operation, the peak target linear power is a critical parameter for assessing the perfonnance of the target decay heat removal system, especially during the transition from normal cooling to backup cooling after a reactor trip. The calculated peak linear power is to a reactor core power at 10 MW 1 and an extreme core bumup case Nevertheless, this value is conservatively used for the entire irradiation period. A worst-case control blade age is assumed (0 and 8 years) and the beryllium reflector age is assumed to be 0 years. This worst-case value conservatively corresponds to a Page 16of190 ATTACHMENT 1 control blade withdrawal position at despite the fact that the "critical position" is The estimated 2cr uncertainty m the peak linear power is -based on the same statistical combination of parameters as for the maximum target power. The peak linear power to be used for safety analyses is -1.8 Codes and Standards for Design, Fabrication and Operations The codes and standards; design methods; software; design documentation; quality assurance and software V & V are listed in Attachment 3 to the License Amendment. 2. Detailed Selective Gas Extraction Target Assembly Description and Structural Analysis This section describes each of the separate subsystems comprising the Target Experimental Facility (TEF): Target Assembly (TA), Target Cooling System (TCS) and target transfer system, including the neutronic and thermal-hydraulic design for the SGE TEF. 2.1 Target Assembly 2.1.1 Functions The functions of the TA are as follows:

Produce Mo-99 by the fission of U-235 -with a nominal enrichment of-but not to exceed 19.9%.

Provide containment, support, and positioning of the

target material for irradiation and cooling.

Guide, direct, and effectively distribute cooling water from the TCS to the surfaces of the clad target rods containing**

Prevent fission products from inadvertently entering the reactor pool water and/or containment building.

Maintain the target pellet configuration such that the TAs are always sub-critical, and ensure that any neutronic coupling between the T As and the MURR reactor core will continue to meet the reactor license Technical Specification (TS) requirements.

Enable detachment of the target cartridge from the target housing at the end of irradiation, for short-tenn transfer to in-pool storage. Page I 7of190 ATTACHMENT 1 2.1.2 Mechanical Design The two T As are designed to be installed in the graphite reflector positions . The TAs are mirror images of each other and differ only in the position of the inlet cooling water pipe. Each TA consists of a water inlet section, a target housing, a lower plenum, a cartridge, an outlet diffuser, and a cartridge locking mechanism. The housing is held laterally in place by an indicator hole in the reflector support plate and vertically in place by the TCS inlet piping which includes a compressible link. The TA components are shown in Figure 4. The modeling and mechanical design of the TA components in 3D was perfonned using the commercially available SolidWorks 2016 software package. Figure 4 Target Assembly 2.1.3 Target Housing The functions of the target housing are to direct the flow of the cooling water, provide structural support, and position the cartridge within the graphite reflector region. The vertical and plan cutaways of the target housing are shown in Figure 5. The target housing is fabricated from welded -plates while the lower housing plenum is fabricated from -* Cooling water is fed to the target housing by the TCS through a line at the top of the assembly. The target housing then directs the water flow down Page 18of190 ATTACHMENT 1 through the lower plenum, up the inside of the cartridge and finally out through the diffuser into the reactor pool. The target housing and the lower housing plenum are bolted together and water leakage is prevented by a metal c-seal that keeps this interface watertight. Figure 5 Illustration of Target Housing Elevation and Plan Views The bottom of the housing has an indicator stub that locates the housing in the reflector support plate, which bears part of the weight of the TA and attached piping. The piping is secured by brackets within the reactor pool and part of the weight load from the piping is transferred to the lower housing plenum by the TA sides. 2.1.4 Target Cartridge The functions of the cartridge are to ( 1) position and support the target rods containing the LEU pellets, (2) provide a cooling passage for the target rods, (3) reduce neutron flux peaking at the center by a neutron absorber, and (4) to mix and guide the water outlet flow. The cartridge consists of an -diffuser on the top, an -cartridge flow housing, at the front and sides of the cartridge flow housing, an -lower cartridge flange, a locking mechanism, and

target rods. The target rods are vertically oriented in a single plane positioned in the target cartridge orthogonal to the direction of the neutron flux (Figure 6). During commissioning operations, it will be necessary to load fewer than -filled target rods in a cartridge. When fewer than -filled target rods are loaded, the remaining positions are filled with Page 19of190 ATTACHMENT 1 stainless steel filler rods with the same dimensions to ensure that the flow around all rods is uniform and at design conditions. The orientation and position of the cartridge relative to the reactor maximizes the Mo-99 production while meeting temperature limits on the target rods. The target rods are held in position by the top and bottom cartridge flanges which allow for the coolant to flow around the target rods. The cartridge is secured in place by a locking mechanism located on top of the diffuser. The locking mechanism engages and disengages to the top of the target housing. Figure 6 Cartridge Configuration and Target System Section View Figure 7 shows the cartridge upper and lower sections highlighting key components and features of the design. The cartridge design is a clamshell with two seam welds running the length of the cartridge. Five (5) pairs of -pins ( are located along the center axis of the clamshell assembly to prevent excessive stresses at the weld locations due to the higher internal pressure from cooling water flow. This simplifies the fabrication process as well as allowing more control over the tolerances for the fit between the water cooling channels and the target rods. The cartridge is first located to the target housing by a pair of guide rails that lead to a pair of locating pins. These high tolerance pins are part of the target housing lower plenum and receive and place the cartridge in its final position in relation to the reactor. The cartridge is then held in place by a pair of locking features in the upper TA locking mechanism. The features are actuated via a lever that locks and unlocks the cartridge. Page 20 of 190 ATTACHMENT 1 Figure 7 Cartridge Upper and Lower Sections The target rods are fixed on the top (upper endcap) and laterally supported at the bottom (lower endcap) The target rods are located and held concentric to the water channels by-features that are fabricated into the lower cartridge flange and can be seen in Figure 8. This ensures even flow velocities around the target rods through the cartridge. The water cutouts on the cup features are smaller than the pointed end cap of the target rod. This eliminates the possibility of the target rod getting stuck on one of these water bypass features when inserting them into the cartridge and ensures the operator can always find the center of the cartridge and guide the target rod into its final position. Page 21 of 190 ATTACHMENT 1 Figure 8 Target Rod Lower End Cap Pins Position Rods Relative to Lower Housing Water Plenum The water cutouts are designed to create near-uniform coolant flow over the target rods as soon as the water enters the cartridge. 2.1.5 Target Rods Figure 9 shows an individual target rod assembly, which consists of an upper end cap, cladding, a spring, -target pellets and a lower end cap. The end caps are fabricated from -bar with integrated features designed to optimize installation and extraction from the cartridge. Both end caps are welded to the cladding autogenously (no weld rod) by a standard The individual pellet/clad components and dimensions are listed in Table 3. The cladding will be fabricated and inspected in accordance with seamless alloy tubes for nuclear reactor fuel cladding applications per American Society for Testing and Materials (ASTM) B81 l. The end caps and -spring will be fabricated from bar material in accordance to ASTM B351. Page 22 of 190 ATTACHMENT 1 The filler rods have external dimensions that are identical to the target rods, except that they are fabricated from The filler rods ensure that the required flow conditions for Figure 9 Target Rod Arrangement Table 3 Target and Filler Rod Dimensions (cold) Nominal Value in Inches (mm) Component --Active target rod length (cold) --Total target rod length --Pellet height -* Pellet outside diameter -* Clad ID -* Page 23 of 190 ATTACHMENT 1 2.1.6 Neutron Absorber Figure 10 Neutron Absorber Section View 2.1. 7 Diffuser The functions of the diffuser are to provide upper support and containment of the target rods, provide water mixing for a bulk outlet water temperature measurement, direct the cooling water exiting the targets away from MURR equipment, and to guide nitrogen-16 (N-16) flow away from the reactor pool surface. The water is mixed in the diffuser's flow mixing zone. The exit temperature measurement is used to determine the power of the TA. 2.1.8 Target Assembly Cooling Water Flow Path The TA cooling water flow path is shown in Figure 11. Cooling water enters the housing from the inlet pipe and flows into the open lower plenum turning into the lower cartridge flange. The lower cartridge flange has -features to minimize water bypassing the target rods during nonnal operations. The flow then travels along the target rods, into the diffuser and is ultimately rejected to the reactor pool. Page 24 of 190 ATTACHMENT 1 Figure 11 Inlet and Outlet Plenums with Method of Attachment (water flow is shown in blue) 2.1.9 Upper Target Housing and Cartridge The upper target housing, cartridge, and lower section of the diffuser are shown in Figure 12. The lower flange of the diffuser acts as the lid that holds down the target rods and keeps them secured in the cartridge by capturing the target rod's upper end cap. This, along with the lower cartridge flange supporting the target rods, properly constrains the target rods through the installation, irradiation, and transfer to the temporary loading/unloading station in the reactor pool. The lower diffuser flange is welded to the neck portion of the diffuser, which collects the water exiting the cartridge and mixes it before guiding it to the resistance temperature detector (RTD) for measurement of the outlet temperature, and ultimately discharges it to the pool. The mixing of the flow is significant for accurate measurement of the power being generated by the TA. The lower diffuser flange is also designed such that the water cutouts allow the flow to move through the sections have the same cross sectional surface area as the rest of the conical sections of the diffuser all the way to the exit tee. This allows the diffuser to have a minimal pressure drop from the cartridge exit to the reactor pool. Page 25 of 190

ATTACHMENT 1 Figure 12 Upper Target Cartridge Arrangement (water flow is shown in blue) When it is time to remove an irradiated cartridge from its target housing for transfer to the in-pool loading/unloading station, the cartridge will be maneuvered and handled remotely in the reactor pool by an operator positioned on top of the pool, using the tools designed for cartridge handling. The cartridge is moved from the TA to the in-pool unloading station, where the diffuser is unlatched from the cartridge to access the target rods for loading into the in-pool storage location. The diffuser is attached to the cartridge by four spring flexures that engage the cartridge upper flange. The flexures are engaged and disengaged by depressing a spring loaded lever located at the top of the diffuser. Tools have been developed to perform such a function within the reactor pool. 2.1.10 Target Pellets Manufacturing of the target pellets, the loading and welding into the Page 26 of 190 ATTACHMENT 1 cladding shall be performed in accordance with the required fabrication, inspection and quality control procedures. Figure 13 Target Pellet Geometry 2.1.11 Design Basis for Target System Materials The considerations upon which target system material selections are based are as follows:

Prior operating experience for target cladding materials m a nuclear reactor irradiation environment is preferable.

The cladding materials must have good mechanical strength at both normal operating temperatures and at expected temperatures during transients.

Materials in the neutron flux must have low neutron absorption cross-sections so as not to impede the rate of Mo-99 production.

Selected alloys must have the properties to readily support fabrication into required shapes and must be readily weldable. Page 27 of 190 ATTACHMENT 1

The material used for the target rod cladding must not undergo undesirable chemical reactions with the target pellet material or fission products within the operating temperature range. 2.1.12 NRC-Approved Fuel Cladding Alloys The target rods have leak-tight cladding that has proven prior experience for this application, per the design requirements. Alloys that have been approved by the U.S. Nuclear Regulatory Commission (NRC) for fuel cladding in power and research reactors in the United States fabricated and inspected in accordance to seamless alloy tubes for nuclear reactor fuel cladding applications per ASTM B8 l 1. 2.2 Structural Analysis of Target Assembly 2.2.1 General The TA components are analyzed in accordance with ASME Section VIII Division I and II. The unirradiated allowable stresses used in the analyses for SS316L and Al6061-T6 were obtained from the ASME Section II part D, and are shown in Figure 14. The Zircaloy-4 cladding has been analyzed structurally for both normal and off-normal operating conditions. 2.2.2 Target Housing and Cartridge The structural analysis for the TA was performed using ANSYS 2016. The calculations confirmed that the target housing and cartridge components are properly sized according to the ASME B&PV Section VIII, Division 1 and Section II-D, for normal and off-normal conditions. Based on this analysis, the structural design life of the target housing is conservatively estimated at -while the design life of the cartridge is conservatively estimated for-* bending stresses of show that both the target housing and cartridge aluminum components are well within the allowable limit of8 (Figure 14). At these stresses, this design ensures that both Page 28 of 190 ATTACHMENT 1 the housing and the cartridge maintain a Factor of Safety (FOS) > 3 for the aluminum components and a FOS > 2 for the stainless steel components before the material begins yielding (241 MPa yield for aluminum and 172 MP a for SS3 l 6L per the ASME Code). This satisfies Section C(l)(c)(3) of Regulatory Guide 2.2, "Development of Technical Specifications for Experiments in Research Reactors," on mechanical stress effects for materials of construction for reactor experiments, which states that materials of construction and fabrication and assembly techniques utilized in experiments " ... should be so specified and used that assurance is provided that no stress failure can occur at stresses twice those anticipated in the manipulation and conduct of the experiment or twice those which would occur as a result of unintended but credible changes of, or within, the experiment." Both the housing and the cartridge therefore maintain a FOS > 2 to the yielding allowables which provides plenty of margin to failure stresses as mentioned in Regulatory Guide 2.2. 120 100 iii Q. 80 ::! "' Q) 60 :c ; .2 40 <( 20 0 0 ................. ... ******* ....

316l s.5

6061 Al (T6) *** ............ . **************-*-.!!...... . ........... .. .. ....................................................................... . 50 WO °'* *,. * ..... *-.. , ..... ......... 150 200 250 Temperature (°C) Figure 14 300 350 400 450 Allowable Strength vs Temperature for Al 6061T6 and SS316L 2.2.3 Pellet-Clad Interaction 500 Pellet-clad interaction for the target rods have been analyzed using FRAPCON 4.0 to ensure that the target rod performance does not exceed design limiting factors for target pellet melting temperatures and cladding strain cycles. Each target rod consists of pellets that are encapsulated by the -cladding. Page 29 of 190 ATTACHMENT 1 Figure 15 Pellet and Cladding Details .. The reactor nominal operating power is 10 MW" with a reactor shutdown occurring at the end of the three (3) week period. For analysis purposes it is assumed that at the beginning and end of the --period, the reactor would run at 115% (11.5 MW1) power for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The result, an upsurge in fission gas release, internal pressure rise, and additional thennal growth and pellet relocation are experienced, though the duration and frequency are not enough to cause any design limit to be exceeded. The results of the FRAPCON analyses for the above operating scenario are shown in Table 4. Stresses and strains as a result of pellet-clad interaction are highest for the minimal cold gap -* The results show that the primary pressure induced stress is , which is well within the primary stress limit of at temperature (Figure 16). With a factor of safety (FOS) of > 90 on primary stresses, the design meets ASME B&PV Code as well as Regulatory Page 30 of 190 ATTACHMENT 1 Guide 2.2, Section C(l)(c)(3) on mechanical stress effects for materials of construction for reactor experiments (FOS against failure of > 2). Primary stresses are therefore not the driver for cladding failure. Secondary stresses as a result of thennal differential expansion and re-location of the cracked pellet on the other hand are the main drivers for cladding longevity. A yield strain for -at temperature of-is well within the strain range of twice of yield, 1.6%. Therefore, this meets the secondary stress intensity limit. With respect to cyclic fatigue -' the maximum number of cycles that the cladding can sustain is -as shown in Figure 17 (Reference 3), which includes irradiated specimens. With only-expected cycles, the cladding has ample design margin. Gap Size Temperature (°C) Pellet centerline Pellet surface Cladding ID Cladding OD Cladding Average Strain(%) Radial Axial Hoop Gas Pressure (MPa) Cladding Internal Table 4 FRAPCON Results Summary ----* * * * * * * * ------*

Page 31 of 190 --* * * * ---*

ATTACHMENT 1 600 500 *.. ********* ............ . .... .. 400 ********** ................................. i II) * :ii j C( 300 200 100 0 0 100 200 300 400 Temperature ("C) Figure 16 Allowable Stress 1 % strain (ASME; factor of 2 on stress/strain) ii 100.0 .... -....,* ... .. l!! o. ' * ... "%. .... .... , 'j °""' J. D "o .... JD ._ ________ ....... ...,*, ... !'! ..... .. . .... A-----------t...;;,,-=""f' ... svoo 0

0 --.... __ ._:: -.. _ SO-lO sOOOOOOO f'ouou* Uf.1 ..,. ccrcln> _ _ FIO. 7-Cydicplasrlc11n;J,..,,s.,,,,.,.....o.-,q/C'J'/r110tup1""fordadl/U,,11'kslMfirrodiol or lrm4ill1ttl '""' eyd,. llt ..., EDF' l'WR: -Wlf11Ud sp<<i-u ....,J ....,""""' ttJJ<!ll ar o a, 1uu1 llurlJot f.11' *" tt11H.,

a, > "*"""Mp ro fVllllllY* Figure 17 Fatigue Chart 500 400000 cycles (ASME 20,000 cycles; factor of 20) 70000 cycles I The reported values for cycles in Figure 4-14 were adjusted in accordance to ASME B&PV Code to provide a factor of 2 on stress/strain and 20 on cycles (whichever is more conservative). With a factor of> 2 on stress/strain, ASME B&PV covers the requirements ofRegulato Guide 2.2. This means that in case of I in a factor of 2, or strain, the uanti icles are estimated at . Note that these are adjustment factors to the experimental data set to obtain estimates of lives of components per NUREG/CR-6815. Page 32 of 190 ATTACHMENT 1 In case of the maximum gap -* the upper limit of manufacturing tolerances, the concern is the centerline temperature of the target pellet. The maximum target pellet centerline temperature is --(Table 4), which will occur at startup when no relocation has occurred. This value is well within the pellet melting temperature of The analyses show that the target rod meets the design requirements and can be safely operated under nominal reactor conditions of 10 MWi, including the two (2) 11.5 MW1 excursions at the beginning and end of the three (3) weeks. The design incorporates a safety factor of .. for cladding cycles and I -for pressure-induced primary stress cycles. Additionally, the pellet cladding exhibits a thennal margin of safety of , while the contained target pellet exhibits a thermal margin of safety of 2.2.4 Effect of Neutron Irradiation on Target Assembly Structural Materials The irradiation damage to TA materials of construction was calculated for full neutron spectrum using damage cross-sections (Reference 4). The maximum damage occurs at the active target vertical mid-plane. The driver neutron flux from the reactor core drops off rapidly as a function of vertical distance from the mid-plane. Table 5 shows the component, location and damage in tenns of displacements per atom ( dpa). Table 5 Maximum Irradiation Damage to Target Cartridge Materials Fast Neutron dpa for Design Life (0.1-20 MeV) Material Component Location (weeks) Fluence Design Life (n/cm2) (Ref. 5) -Target Cartridge Front face at active * -* target mid-plane Target Rod Cladding Target rods I --Target Cartridge Front face at active * -* target mid-plane -Target Housing Front face at active * -* target mid-plane Lower Plenum Front face * -* The impact of these low levels of irradiation damage on the material structural properties are as follows: --This alloy has a high degree of radiation tolerance and can sustain damage levels in excess of 100 dpa. The yield and ultimate strengths are increased by above 0.5 dpa. The alloy Page 33of190 ATTACHMENT 1 retains good ductility(> 5% failure strain) after irradiation; however, the AL6061 -strain is very low at the maximum damage location. The use of non-irradiated properties per the ASME Section II code is appropriate for the use of this alloy in the TA (Reference 6). is used reliably 1022 n/cm2 at 300 to 400 °C (572 to 752 °F). for L WRs with fast fluences in the range of . However, the effect of irradiation at lower temperature is to lower the rate of annealing. This increases the rate of damage and the level at which thennal-mechanical properties saturate as a function of fluence (or dpa). Reference 7 shows that at L WR temperatures, the properties tend to saturate at a fast neutron fluence of 1 x 1021 n/cm2, but at research reactor temperatures (< 100 °C), some properties will saturate at , which is lower than the maximum fast fluence experienced by the target. Nevertheless, the total damage to the is not significant for its intended service. The effect on thennal conductivity is negligible (Reference 8). At the lower target operating temperatures, the tensile strength will increase slightly (+20%) at the expense of some loss of ductility (-30%). At --' the irradiation effects will be negligible. These factors are accounted for in the design and structural analysis --This alloy will experience a small increase in tensile strength and a corresponding small loss of ductility for the maximum dose of -at the operating temperature range. This maximum exposure is very small and occurs at the longitudinal center of the cladding. The yield strength increases by about 25% while the failure strain decreases by about 30% (Reference 9). Because the --experiences very low strain along its length, the loss of ductility is not a concern. The fracture toughness will experience a small (< 15%) decrease (Reference 10). This is also an insignificant impact on structural design due to the small negative effect of the low dpa. Therefore, the use of non-irradiated properties per the ASME Section II code is appropriate for the use of this alloy in the TA. 2.2.5 Bowing of Target Rods The LEU target rods are driven by a directional neutron flux from the reactor core; hence, they can be expected to experience a power gradient across the

pellet. This will cause a small degree of bowing of the target rods in the radial plane of the core as a result of the following two (2) effects: 1. The reactor core facing side of the rod will have a slightly higher temperature than the opposite side causing differential thennal expansion; and 2. The reactor core facing side of the rod will be exposed to a slightly higher fast neutron flux causing a small degree of differential irradiation induced strain. Conservative evaluations of both effects have been carried out and the estimated bowing is shown to have no effect on rod heat removal. Page 34 of 190 ATTACHMENT 1 The power density distribution was analyzed for the target rod where the power density of the front half of the pellets was found to be on average

greater than the rear half. At the worst axial location, the front half power density was found to be

greater than the rear half. This worst-case assumption is extremely conservative, as only a few pellets throughout both targets come close to this power skew, and the average skew for rods containing those points is much lower. In addition, the entire stack was conservatively assumed to have the maximum power density of -* The combination of maximum tilt with maximum power density produces a front-to-back temperature difference of about -* A thermal analysis was performed for the worst power density, showing a total axial growth of the target rod cladding of approximately -* Output from a FRAPCON analysis showed that the induced axial growth of the target rod cladding was -* Therefore, it was deemed that the effects of the irradiation induced growth could be captured conservatively in a structural model by doubling the -coefficient of thermal expansion in the structural model. Figure 18 provides a visual depiction of deflection of a target rod for a conservative, worst-case power skew. The figure shows the black wireframe of the un-defonned geometry, with the defonned body Figure 18 Deflection Due to Target Rod Bowing for Worst-Case Front-to-Back Power Skew Given that the distance between the target rod and the cartridge wall when centered is about --no contact would occur. Additionally, flow through the cartridge is highly turbulent and boundary layers are very thin, so this level of deflection would not affect the heat transfer to the coolant. A more realistic average power skew scenario predicts a deflection of Page 35 of 190 ATTACHMENT 1 Given these results, it can be concluded that target rod bowing due to both thennal and irradiation effects will not affect the thennal-mechanical perfonnance of the target rods and cartridge assembly. 3. Target Cooling System The Target Cooling System (TCS) provides cooling water flow to the two (2) T As during nonnal operation and during both planned and unplanned shutdowns. The TCS has two methods of cooling, corresponding to force cooling (pumped) flow during nonnal operation and natural circulation flow during post loss of forced cooling (LOFC) scenarios.

Forced Cooling: The reactor is operating and producing fission power in the TAs. The TCS pumps water from the reactor pool, cools it through a heat exchanger that interfaces with the MURR secondary coolant system. The TCS then transports the cooled water to

T As. The cooling water exits the T As and flows back to the reactor pool.

Natural Convection: In the event that pumped flow to the TAs is interrupted, the reactor is shut down and the TA power reverts to decay heat. The TCS transitions to natural circulation flow to the TAs as driven by natural circulation from target decay heat. The TCS is divided into two subsystems, the target flow assembly and the water cooling module. The flow interface begins at the reactor pool and ends at the TA inlet. The heat rejection interface is with the MURR secondary coolant system for forced cooling operation and with the reactor pool for natural convection operation. The TCS instrumentation also includes inputs to the MURR reactor safety system to scram the reactor if forced flow falls , and on high coolant temperature to the TA. 3.1 Functions The primary function of the TCS is to provide cooling water to the T As at a sufficient mass flow rate to maintain the target material within its required temperature range for safe operation. The system is designed to perform the following functions: 3.1.1 Process Functions

Transfer and reject the target heat to the MURR secondary coolant system during normal operation.

Maintain sufficient cooling water flow rate and temperature conditions through the target to maintain a large CHFR margin and such that any subcooled nucleate boiling has a negligible impact on target rod integrity.

Minimize net exchange of heat between the TCS and the reactor pool water during normal operation. Page 36of190 ATTACHMENT 1

Provide for reliable transition of target heat removal to the reactor pool via natural circulation in the event of an LOFC.

Provide instrumentation to assure that flow, temperatures and pressures are within specified conditions for operation. 3.1.2 Safety Functions

Provide a signal to the MURR reactor safety system to enable a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation* or for high heat exchanger outlet temperature.

Provide a corresponding signal to open the decay heat removal valves to initiate natural circulation flow to the T As that coincides with a reactor scram in the event that the forced cooling water flow rate falls outside specified ranges for safe operation. Accident analysis shows that even if no natural circulation occurs after an LOFC event, target decay heat will be transported directly to the pool. The targets will remain in a safe condition with no radioactivity released to the reactor pool. 3.2 Target Cooling System Principal Design Parameters The principal design condition for the TCS corresponds to both T As loaded with

target rods each and the reactor operating at 10 MW1* The target power levels will vary slightly due to shifts in the driving neutron flux distribution caused by changes in burnup of reactor fuel elements, position of control rods and age of the beryllium reflector. The TCS is designed to accommodate the maximum target thermal power and the maximum linear heat rate. The TCS is designed to maintain an approximate -cooling water temperature rise across the target cartridge(s) for the maximum power condition. The corresponding flow rate combined with the cartridge design assures the target rod CHFR is greater than 2.0 for the worst-case allowable operating condition. Table 6 summarizes the principal TCS design parameters assuming the reactor is operating at 10 MW1 and the combined target loading is 22 target rods. The system is controlled to provide --mass flow which is divided equally . This flow rate is held constant for all operating conditions. It includes an additional

flow to compensate for any leakage that might occur through the labyrinth seal between the cartridge and the target housing. The inlet temperature is adjusted so that the outlet of each TA is at or slightly above the reactor pool temperature, which is nominally Page 37of190 ATTACHMENT 1 Table 6 Target Cooling System Design Parameters for Two Targets 2 Para meter Targ et rods heat generation rate, kWt .. .. Targ et structure heat generation rate kWt ..

Targ et coolant mass flow rate including. leakage bypass, kg/s Targ et System inlet temp, °C Targ et System outlet temp, °C (Reference at nominal flow) 3.3 Target Cooling System Design Theta rget cooling system (TCS) is comprised of two subassemblies:

The Target Flow Assembly

The Water Cooling Module 3.3.1 Target Flow Assembly Design Aproc ess flow diagram of the target flow assembly is shown in Figure 19. The Water Cooling Module (WC rejecti single M) draws water from the reactor pool through the inlet pipe and cools it to the desired value by ng the heat to the MURR secondary coolant system. The cooled water exits the WCM through a pipe that splits into supply lines . Each supply line has a throttle valve that is Uy adjusted to balance the flow between the -or to shutoff flow to the TA. After the manua flow is balanced, the valves are pinned to prevent inadvertent maladjustment or closure. Safety -related flow meters on each supply line provide redundant signals to the MURR reactor safety to enable a reactor scram on high or low flow. RTDs at the TA inlets and the inlet to the WCM d to control the cooling water inlet temperature to the T As. system are use Each supply line has redundant natural circulation bypass valves, which in nonnal operation are atically held closed, but are spring-loaded to automatically open on a low flow signal from the d line. This allows reactor pool water to enter into the supply line to initiate natural circulation. pneum affecte 2 The H eat Exchanger will have a capacity of *********** , of the total (rods + structure) target heat on rate. generati Page 38of190 ATTACHMENT 1 After flowing through the T As, the cooling water is returned to the pool through the diffusers located on pipe extensions from the target cartridges. The purpose of the diffusers is to slow down and redirect the flow to allow decay of any N-16 before it can circulate to the pool surface. The diffusers also serve as the mounting point for the RTD to measure the target outlet temperature to provide a caloric measurement of the target power. Figure 19 Process Flow Diagram of Target Cooling System Figure 20 shows an elevation view of the TCS, while Figure 21 shows the TCS superimposed on the reactor in-pool systems and biological shield. The WCM is located on the third level of the Costar Tower in the reactor containment building. The supply line is split at the wye just before the access bridge between the Costar Tower and the reactor biological shield. The individual supply lines to each TA are run parallel to the access bridge to the reactor pool. Flexible couplings are located on each line that crosses from the Costar Tower to the biological shield to accommodate any differential displacement of the two structures. The pipe to the wye is 4-inch SS316L Schedule 40. Downstream of the wye, the piping is 3-inch SS3 l 6L Schedule 40 up to the pool surface. There the pipe transitions to Al606 l Schedule 40 via a flanged coupling. Page 39of190 ATTACHMENT 1 Figure 20 Elevation View of Target Cooling System Flexible ....--couplings The downcomer supply pipes are supported from the floor structure at the top of the biological shield and the T As are allowed to rest on the reflector support plate. A flexible coupling is located above each TA to accommodate any growth or differential motion between the TA and the pipe upper support. A single suction line draws water from the pool and is delivered the WCM. The other pump is isolated using a ball valve. The suction line is 3-inch Al6061 Schedule 40 in the reactor pool and transitions to SS316L outside of the pool. The water suction pipe inlet is fitted with a replaceable screen to prevent blockage by debris. The decay heat removal valves on each supply downcomer are positioned at a submerged location just above the refueling level. These valves are 2-inch butterfly valves with SS316L bodies. Each valve has an actuator rod that extends to a pneumatic actuator located above the pool surface. Page 40 of 1 90 ATTACHMENT 1 TCS Module MURR Costar Tower Figure 21 Target Cooling System Superimposed onto MURR Reactor Pool and Biological Shield 3.3.2 Water Cooling Module Design Figure 22 shows the process flow diagram for the WCM, which incorporates two (2) redundant and connected coolant pumps and heat exchangers. If a pump or heat exchanger is taken out of service, the installed spare component can be placed in operation to permit continued TEF operation. The pumps are self-priming, centrifugal pumps with variable frequency drives (VFD), which are manually controlled to provide the required flow. The heat exchangers are stainless steel, plate-type units. The heat exchangers reject heat to the MURR secondary coolant system. The temperature of the heat exchanger outlet flow is controlled by a flow bypass valve on the secondary side of the heat exchanger. A target system P&ID drawing is presented in Figure 23. Page 41 of 190 ATTACHMENT 1 Figure 22 Sch.ematic for the Water Cooling Module Page 42 of 190 ATTACHMENT 1 Figure 23 Target Cooling System Page 43of190 ATTACHMENT 1 The WCM is located on the third level of the Costar Tower at approximately the same elevation as the top of the biological shield. The coolant pumps, heat exchangers, piping and control panel are mounted on a single pallet and installed as a single unit. Table 7 describes major specifications of coolant pumps and heat exchangers for the TCS. Table 7 Major Component Specifications ., Units SI English Pump . Nominal Flow Rate(+/- 15%) --Max Head with 25% margin --Outlet Pressure with 25% margin --Max Inlet temperature --Material -* Drive -* NPSH -* Pump Type -* Inlet flange pipe size -* Outlet flange pipe size -* Nominal Power --Voltage ....

Current Normal Operation (460 VAC)

Frequency ..

Phases I

_, Heat Exchanger . Heat duty with 15% margin -* Hot side flow rate (primary) --Hot side inlet temperature nominal (primary) --Hot side outlet temperature nominal (primary) -.. Hot side pressure drop (primary) --Differential approach temperature (for reference) --Cold side flow rate (secondary) --Cold side inlet temperature (secondary) -.. Cold side outlet temperature (for reference) --Maximum allowable working pressure (MA WP) -.. Cold side pressure drop --Material -* Type

Page 44of190 ATTACHMENT 1 3.4 Target Cooling System Performance 3.4.1 Thermal-Fluid Performance in Normal Operation, Mode 1 The TCS must be on Mode 1 prior to and during the start of the reactor if the T As contain active target rods. The process flow diagram with definition of state points for one (of the two) T As is shown in Figure 24. The figure also provides the flow, temperature and pressure corresponding to each state point. Figure 24 State Points within the Target Cooling System 3.4.2 Thermal-Fluid Performance in Natural Circulation Flow, Mode 2 Figure 25 shows a schematic of the TCS operating in the natural circulation mode. In the event of an LOFC, the reactor is tripped by a low flow signal cm of nominal flow) if the event is cause by a reduction of flow (e.g. loss of offsite power) or a high flow signal ( .. of nominal flow) if the event was caused by a pipe break downstream of the sensor. When the reactor is tripped, the driver neutron flux to the targets quickly decays and the target thennal power transitions from fission heat to fission product decay heat. Coincident with the reactor trip, an actuation signal automatically opens the two (2) decay heat removal valves on the affected downcomer pipe(s) such that the downcomer is open to the Page 45 of 190 ATTACHMENT 1 reactor pool. This allows natural circulation of pool water to the TA and back to the pool through the diffuser. Figure 25 *Target Cooling System P&ID Page 46 of 190 ATTACHMENT 1 The analysis of the TCS performance in the natural circulation mode for the set of design basis events that can lead to a LOFC is provided later in this document. In all design basis cases, the natural circulation flow to the TA provides the required cooling. 3.4.3 Structural Performance During a Seismic Event A seismic analysis was perfonned on the piping components of the TCS based on the acceleration and spectra given in the American Society of Civil Engineering Code 7-10 and interpreted according to American Society of Mechanical Engineering B&PV B3 I .3. The specified design spectra corresponds to an acceleration of 0.113 g. The analysis was perfonned using AutoPipe, a specialized nonlinear finite element piping program. The WCM and TA flanged connections where considered idealized anchors. The piping supports are rail-mount vibration damping clamps and are modeled as line supports with no gaps and firmly anchored to the biological shield, the Costar Tower or the bridge. Figure 26 shows the location and type of pipe supports for piping located above the reactor pool. To account for independent seismic movement of the building to the pool, flexible piping is added at the bridge level. All flexible piping is 11 inches in length and flanged. A guide support is added with the flange connections on the non-flexible piping side. Stiffness values were based on General Atomics' test data. Figure 27 shows the location and type of pipe supports for in-pool piping. N y All other supports shown are Guide Supports Only Figure 26 exch*ngers. Rigid Anchor connections on the ends. Flexible Joints to simul*te flexible pipine. (Not shown on Solidworks model) Location and Type of Supports on Above Pool Piping Page 47 of 190 ATTACHMENT 1 From these flanges all piping in the Guide+ Line Supports All other supports shown are Guide Supports Only negative Y direction is aluminum Flexible Joints to simulate flexible piping with anchors at target housing interface Figure 27 Location and Type of Support on In-Pool Piping Two (2) types of analyses were perfonned. The first type considers the biological shield, bridge and Costar Tower as rigid structures due to the low seismic loading. The maximum deflections of structures within areas of interest are estimated to be < 1/161" inch. Table 8 shows the key results of the rigid structure analysis. The table shows that the minimum safety factor is 7.8 at the worst location. Page 48 of 190 ATTACHMENT 1 Table 8 Results of Rigid Structure Seismic Analysis Location Material Stress Stress Allowable Safety Factor (ksi) (ksi) Y-pipe, before flexible piping SS316L 0.7 16.7 24 Inside pool, on expansion loop 6061-T6 1.0 16.7 17 At heat exchanger interface SS316L 3.2 25 7.8 At SS to Al pipe interface 6061-T6 0.2 25.1 126 3-inch to 2-inch Y-pipe SS316L 0.3 16.7 56 All Al piping 6061-T6 0.3 16.7 56 Y-piping support SS316L 0.9 22.2 25 Inside pool, on expansion loop 6061-T6 1.0 22.2 22 The second type assumes that the biological shield/pool and Costar Tower combined with the bridge will laterally displace during a seismic event. To demonstrate the effectiveness of the flexible couplings, a hypothetical forced displacement of 2 inches is applied. By comparison, an initial seismic evaluation of the Costar Tower showed a less than 1/8111 inch lateral movement at the location where the bridge connects. The AutoPipe model showed that the displacement has negligible effects on the stress ratio (max ratio is 0.05) demonstrating that the flexible piping design works as intended. 3.4.4 Secondary Cooling for the Target Cooling System The heat from the TCS is transferred to the MURR secondary coolant system by means of one of the TCS heat exchangers. The heat is then dissipated to the atmosphere through a cooling tower with seasonal additional cooling provided by a chill water system. The secondary coolant loop that serves the TCS consists of one (1) coolant circulation pump, two (2) heat exchangers, two (2) automatic temperature control valves, one (1) chill water heat exchanger, temperature and flow instrumentation, and associated valves and piping. A piping and instrumentation diagram is shown in Figure 28. The secondary coolant loop that serves the TCS shares a cooling tower, radiation monitoring instrumentation, and chemical control systems with the existing 15 MW capacity MURR secondary coolant system. Only the secondary coolant loop serving the TCS will be described in this section. The secondary coolant system is designed such that a failure or malfunction of any system component will not lead to target rod damage or an uncontrolled release of radioactivity to the environment. 3.4.5 Circulation Pump The secondary coolant circulation pump for the TCS is a centrifugal, single-stage pump that is connected to a variable speed drive (VSD) unit through a coupling. This pump, designated SP-5, is capable of supplying 500 gpm at a discharge pressure of 50 psig. Three (3) other secondary coolant Page 49 of 190 ATTACHMENT 1 pumps are installed in a parallel configuration with each pump capable of supplying 2,200 gpm to the secondary coolant system. In case pump SP-5 failed, TCS secondary coolant flow can be established using a cross-connect pipe to the discharge header of the other three (3) secondary coolant pumps. Additional detail of the SP-5 control system is provided in Instrumentation and Control section of this document. en c:.. ::j I ... "' Cl OCll 0 c: Q "C r--c z!( ;io 'fl. _. Cll c:.. Q (I) w ;!i! I ..... 'P ... "' ... en c'.i N i Figure 28 Secondary Cooling for Target Cooling System Page 50 of 190 Cll z 0 :c [ij 0 0 z rn rn "" en c:.. 8l ATTACHMENT 1 3.4.6 Heat Exchangers The secondary coolant flow makes a single pass on one of the TCS heat exchangers. The TCS heat exchangers are plate-type heat exchangers. On one side of the plates is secondary coolant, and the other side has TCS coolant. Only one TCS heat exchanger is required for TCS operation. The other TCS heat exchanger is an installed spare for operational reliability. During the summer months, additional cooling to the secondary coolant supply to the TCS may be needed to ensure a sufficient secondary cooling water temperature to maintain proper TCS heat exchanger outlet temperature. This additional cooling to the secondary coolant system is provided by a chill water system via a plate-type heat exchanger. 3.4.7 Automatic Temperature Control Valves The amount of secondary coolant passing through the TCS heat exchanger is controlled by an automatic temperature control valve located in a bypass line of its associated heat exchanger. The valve is an electro-mechanical butterfly valve which responds to the TCS heat exchanger outlet temperature which maintains a constant cold leg temperature to the T As. Since only one TCS heat exchanger is operated at any time, only one automatic temperature control valve is in operation at any time. The other valve is an installed spare for operational reliability. The automatic temperature control valves for the TCS heat exchangers are designated as S-3A and S-3B. Additional detail on the operation of the automatic temperature control valves and system is provided in Instrumentation and Control section of this document. 3.4.8 Instrumentation The secondary coolant inlet and outlet temperatures to both the chill water and TCS heat exchangers and the total flow in the secondary system TCS loop are displayed and recorded in the reactor control room. 3.4.9 Electrical Power System Supporting the TCS Electrical power to the TCS pumps and their Instrumentation and Controls Uninterruptible Power Supply (I&C UPS) will be supplied by Motor Control Center No. 4 (MCC-4) through a new 480/277 V distribution panel. Nonnal Electrical Power System Substation B supplies 480-volt, 3-phase, 60-cycle electrical power to MCC-4 through a 250-amp breaker as described in Chapter 8 of the MURR Safety Analysis Report (Reference 11). The new 480/277 V, 200-amp panel, designated HVP-4, will be connected to MCC-4 in the reactor containment building. Electrical power to the TCS pumps will be provided by HVP-4. HVP-4 will also supply electrical power to a new 120/208-V Distribution Center, designated 120/208V Distribution Center 3, through a 112.5-kVA transfonner. A Lighting Panel on 120/208V Distribution Center 3 will supply electrical power to the TCS pumps I&C UPS as shown in Figure 29. Page 51of190 ATTACHMENT 1 Location of the new electrical wiring will ensure that no electromagnetic interference will exist between the electrical power service and any safety-related I&C circuits. 1 I COLLING 3 PUMP 5 7 9 11 13 15 17 19 21 23 25 27 29 Frol'l MCC 114 Breoker A6 I HVP-4 480/277 200A PUMP 2 4 . 6 480-277V 8 : 10 12 ' 14 112 KVA 16 3PH 18 20 22 24 I 1 26 I TCS 18.C 3 28 UPS Panel 30 5 7 9 Figure 29 120/208V DISTRIBUTION CENTER-3 400A 2 4 6 8 10 Electric Power Supply to TCS Pumps and l&C UPS 4. Instrumentation and Control System This section describes the operating characteristics of the TEF Instrumentation and Control (I&C) Systems. These systems assure that the TEF can be safely operated, monitored, and shut down as warranted. 4.1 Summary Description The TEF I&C Systems are comprised of the sensors, electronic circuitry, displays, and actuating devices available to provide the infonnation and means to safely control the TEF and avoid or mitigate potential accidents that could affect the TEF or the reactor. The TEF I&C systems include the following:

Target Cooling System (TCS) Control System

TCS Protection System Page 52 of 190 ATTACHMENT 1

TCS Parameter Indication, Recording, and Alann System

TCS Secondary Coolant Control System

TCS Secondary Coolant Parameter Indication, Recording, and Alarm System

N-16 Reactor Power Monitoring System

Pool Coolant Monitoring System The TCS Control System is designed to control TCS pump speed and the position of the Decay Heat Removal System Automatic Valves. It is a hardwired system that relies on relays, controllers, push buttons, switches, and light indicators. The TCS Control System receives inputs manually and from the TCS Protection System. The TCS Protection System is designed to prevent operation of the TEF in regions in which target rod damage could occur. This is accomplished by the TCS Protection System which can initiate a reactor scram via the reactor safety system, which instantaneously drops the reactor control rods and takes the reactor subcritical. Therefore, the power of both the TEF and the reactor are quickly reduced. Inputs which govern the TCS Protection System output are supplied by TCS flow and temperature transmitters. The TCS Parameter Indication, Recording, and Alarm System is designed to provide control room operators with indications of operating parameters for the TCS, to record TCS parameter data for term retention or review, and to actuate alarms to alert the operator to abnonnal parameters. This is accomplished by parameter current loop signals going to chart recorders which also serve as parameter indication and alann units. Parameter inputs into the TCS Parameter Indication, Recording, and Alarm System are multiple TCS temperature, pressure, and flow signals. The TCS Secondary Coolant Control System is designed to control the TCS secondary coolant circulation pump (SP-5) speed and the position of the operating target cooling automatic temperature control valve (S-3A or S-3B). It is a hardwired system that relies on relays, controllers, push buttons, switches, and light indicators. The TCS Secondary Coolant Control System receives inputs manually and from the TCS Parameter Indication, Recording, and Alarm System. The TCS Secondary Coolant Parameter Indication, Recording, and Alann System is designed to provide control room operators with indication of parameters convenient for operating the TCS Secondary Coolant Control System, to record TCS secondary coolant parameter data for long-term retention or review, and to actuate alanns to alert the operator to abnormal parameters. This is accomplished by parameter current loop signals going to chart recorders which also serve as parameter indication and alarm units. Parameter inputs into the TCS Secondary Coolant Parameter Indication, Recording, and Alarm System are multiple TCS Secondary Coolant Control System temperature, flow, and valve position signals. The N-16 Power Monitoring System (N-16 PMS) is designed to provide control room operators with indication of reactor core power by measuring the amount of N-16 produced in the primary coolant system. Reactor fission power is directly proportional to the amount ofN-16 atoms produced by the fast Page 53 of 190 ATTACHMENT 1 neutron flux in the reactor. Therefore, reactor core power can be measured separate from the TEF power. The N-16 PMS has two (2) reductant detectors located in mechanical equipment room 114 and two (2) redundant displays located in the control room. The Pool Coolant Monitoring System (PCMS) 1s designed to provide control room operators with indication and an a Jann of elevated iodine-131 (1-131) activity levels in the pool coolant system. Elevated 1-131 activity levels in the pool coolant system would be an early indication of a leaking target rod in the reactor pool. The PCMS has a detector, display, and alanns in the entryway to mechanical equipment room 1 I 4 and has an indication and alann in the control room. 4.2 Target Cooling System Control System Description The TCS Control System is designed to control TCS pump operation and the position of the Target Decay Heat Removal Valves (TDHRVs). From the TCS Control System, the operator can start, stop, and change speed of the TCS pumps. In addition, the TCS Control System contains an interlock between the TCS pumps and the TDHRV position that secures the TCS pumps when any one of the TDHRVs open. The interlock is actuated from a position indication switch located on each TDHRV actuator which opens a contact de-energizing both of the TCS pump control circuits. The control system architecture for the TCS is shown in Figure 30. A larger image of the TCS Control Panel is shown in Figure 31. MURR CONTROL ROOM Control Panel with Hardware Push Buttons/ Switches and Indicators Process Instrument Independent Hard Wired Safety System Functions: -Hardwired isolated interface with MURR SCRAM loop MURR CONTROL SYSTEM ._ _____ _ MURR SCRAM LOOP t Flow, 4*20mA Temperature, 4-20mA VFDs with accessible control pads for local maintenance control COOLING WATER MODULE Figure 30 Critical Process Instrument Target Cooling System Control System Architecture Page 54 of 190 ATTACHMENT 1 When the TDHRVs are in 'Manual' mode, the operator can individually change the position of each TDHRV as needed for maintenance or testing. In 'Auto' mode, low or high TCS flow will automatically open the two (2) TDHRVs in the associated TCS branch of the abnormal flow condition. When a low or high TCS flow condition occurs in a TCS branch, one or more contacts in the TCS Protection System will open and de-energize the two (2) solenoid-operated valves that are applying air pressure to the TDHRV actuators in that TCS branch. Air pressure will be vented off the TDHRV actuators and spring force will rapidly open the TDHRVs, when required. Figure 31 Target Cooling System Control Panel 4.3 Target Cooling System Protection System Description The TCS Protection System initiates a reactor scram via the reactor safety system to prevent operation of the TEP in regions in which target rod damage could occur. The TCS Protection System will initiate a reactor scram based on low TCS flow rate -normal flow), high TCS flow normal Page 55of190 ATTACHMENT 1 flow), and high TCS heat exchanger outlet temperature (-). These set points are based on the limiting conditions used in the thennal-hydraulic safety analysis for the TEP. The basis for the TCS low flow scram is to ensure target fission heat generation is terminated upon a low flow condition caused by a TCP failure, TCS piping break upstream of the flow transmitters or on the TCP suction, or TCS piping blockage. The basis for the TCS high flow scram is to ensure target fission heat generation is terminated upon a TCS pipe break between the flow measuring element and a TA which would reduce flow resistance and increase flow indication at the associated flow transmitter. This condition causes low flow to the TAs. The basis for the TCS high temperature scram is to ensure target fission heat generation is terminated . upon a high temperature condition that is not associated with a change in TCS flow rate. Insufficient secondary coolant flow and high secondary coolant temperature could cause such a high TCS temperature to occur. The TCS Protection System is designed with all analog components. The components and system are designed to be fail-safe, meaning that any loss of power or signal will cause a reactor scram. In addition, all TCS Protection System signals are separate from TCS Control System signals. Keyed bypass switches to the TCS Protection System allows for reactor operation with the TCS not operating, one TCS branch operating, or both TCS branches operating. The keyed bypass removes the protection signal inputs for any part of the TCS that is not operating. 4.3.1 Target Cooling System Low and High Flow Rate Protection TCS flow is measured at the following locations with the indicated TCS Flow Elements (TCFEs) and TCS Flow Transmitters (TCFTs): Flow elements TCFE-lA and TCFE-lB for Target A and Target B, respectively, are located in each TCS branch. The differential pressure caused by TCS flow is measured by flow transmitters TCFT-IA and TCFT-2A for TCS branch A and TCFT-lB and TCFT-2B for TCS branch B. The output signal (4 to 20 mA) generated by each flow transmitter is directed to a square root converter which provides a linear output signal for the two dual alarm trip units and a chart recorder in series with the signal. In addition to providing flow indication and recording, the recorder will initiate a "TCS Lo Flow" alarm when TCS branch flow to either target decreases to 90% of its normal value. If TCS branch flow decreases to 85% normal flow, then the following actions are initiated by the system: 1. Reactor scram; 2. Opening of the two (2) TDHRVs in the associated branch; and Page 56of190 ATTACHMENT 1 3. Actuation of the "TCS Flow Scram" annunciator alann. The low flow dual alarm unit for TCFT-lA opens two (2) contacts: 1. One contact opens in the "Yellow Leg" of the TCS Protection System which in tum opens contact TCS-1 in the "Yellow Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The low flow dual alann unit for TCFT-2A opens two (2) contacts: 1. One contact opens in the "Green Leg" of the TCS Protection System which in tum opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-lB opens two (2) contacts: I. One contact opens in the "Yellow Leg" of the TCS Protection System which in tum opens the contact TCS-1 in the "Yellow Leg" of the reactor safety system which causes the scram (See Figure 32). 2. Another contact opens m the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The low flow dual alarm unit for TCFT-2B opens two (2) contacts: I. One contact opens in the "Green Leg" of the TCS Protection System which in tum opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. In addition to a low flow condition, the TCS Protection System will provide protection on TCS high flow rate. In this case, the chart recorder will initiate a "TCS Hi Flow" alarm when TCS branch flow to either target increases to .. of its normal value. IfTCS branch flow increases to .. normal flow, then the following actions are initiated by the system: Page 57 of 190 ATTACHMENT 1 1. Reactor scram; 2. Opening of the two (2) TDHRVs in the associated branch; and 3. Actuation of the "TCS Flow Scram" annunciator alarm. The high flow dual alann unit for TCFT-lA opens two (2) contacts: 1. One contact opens in the "Yellow Leg" of the TCS Protection System which in turn opens contact TCS-1 in the "Yellow Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens m the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The high flow dual alarm unit for TCFT-2A opens (2) two contacts: 1. One contact opens in the "Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lA and TDHRV-2A actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The high flow dual alarm unit for TCFT-lB opens two (2) contacts: 1. One contact opens in the "Yellow Leg" of the TCS Protection System which in turn opens contact TCS-1 in the "Yellow Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. The high flow dual alann unit for TCFT-2B opens (2) two contacts: 1. One contact opens in the "Green Leg" of the TCS Protection System which in turn opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor scram (See Figure 32). 2. Another contact opens in the TCS Control System that de-energizes both solenoid-operated valves applying air pressure to TDHRV-lB and TDHRV-2B actuators which allows these valves to open. The opening of this contact also activates the "TCS Flow Scram" annunciator alarm. Page 58of190 ATTACHMENT 1 4.3.2 Target Cooling System High Temperature Protection TCS combined cold leg header temperature is measured for use in the TCS Protection System at the following location with the indicated TCS Temperature Elements (TCTEs) and TCS Temperature Transmitters (TCTTs):

Downstream of the TCS Module on the TCS combined cold leg header Resistance Temperature Detectors (RTDs) TCTE-2 and TCTE-3 are located beside each other in the TCS combined cold leg header. TCS Temperature Transmitters TCTT-2 and TCTT-3 convert the associated RTD signal into 4 to 20 mA output signal providing a linear output temperature signal for a dual alarm trip unit and a chart recorder in series with the signal. In addition to providing TCS cold leg temperature indication and recording, the chart recorder will initiate a "TCS Hi Temp" alarm when TCS cold leg temperature increases to -* If TCS cold leg temperature increases to -' a reactor scram and a "TCS Temp Scram" annunciator alarm are initiated. The high temperature dual alann unit for TCTT-2 opens a contact in the "Yellow Leg" of the TCS Protection System which in tum opens contact TCS-1 in the "Yellow Leg" of the reactor safety system which causes the scram. The high temperature dual alarm unit for TCTT-3 opens a contact in the "Green Leg" of the TCS Protection System which in tum opens contact TCS-2 in the "Green Leg" of the reactor safety system which causes the reactor to scram (see Figure 32). Page 59 of 190 ATTACHMENT 1 Figure 32 Target Cooling System Protection System Relay Inputs to the MURR Reactor Safety System Page 60 of 190 ATTACHMENT 1 4.3.3 Target Cooling System Protection Bypass Capability When it is desired to operate the reactor without operating the T As, -bypass keys will be inserted and switches will be placed in the bypass position for the associated TA's (or assemblies') protection to be bypassed. The bypass switches will be a four-position, standardized type switch designed 1 S30 and 1 S31. With no key in bypass switch 1S30 or 1S31, the switch is locked in the Normal" position and all TCS protection signals can affect a reactor scram. In addition, the TDHRV to TCS pump interlock is active to shut off the TCS pump when any TDHRV opens. When a key is in bypass switch 1 S30, the switch is able to be positioned in the *-Bypass,' *-position. With 1830 in either the or the --position, the "Yellow Leg" TCS flow protection scram for the bypassed TCS branch is disabled. With 1 S30 in either the position, the "Yellow Leg" TCS flow protection scram and the "Yellow Leg" TCS heat exchanger outlet high temperature scram for the bypassed TCS branch are disabled. However, bypass switch 1830 only disables the "Yellow Leg" protection. Bypass switch 1 S31 has the same functions as 1 S30, but it only disables the "Green Leg" protection. Therefore, to bypass scram protection from one or more TAs and remove the associated TDHRV to TCS pump interlock, both switches must have their keys installed and positioned correctly. This design ensures that no single switch or single human error will inadvertently remove all scram protection or the TDHRV to TCS pump interlock to a TA that is being irradiated. For example, if only Target I is to be irradiated, the TCS would be lined up to provide cooling flow to only TCS branch 1-TCS branch I would have no flow and the TDHRVs in TCS branch I would be open. Therefore, two keys are inserted and switches 1 S30 and 1 S31 would be positioned to 'Target I Bypass' so that the no flow condition in branch I would not cause a scram and the open TDHRVs in branch I would not prevent running either TCS pump. When switches 1 S30 and 1 S31 are in 'Target I Bypass,' relays are energized that close contacts that bypass the reactor safety system contacts actuated by TCFTll and TCFT. in both legs of the TC8 Protection System (See Figure 32). In addition, these energized relays close contacts that bypass the TDHRV. and TDHRV. valve contacts in the TCS pump controllers thereby allowing the TCS pump to run while the cooling branch I TDHRVs are open (See Figure 33). If is to be irradiated, the TCS pumps would both be secured, and all four TDHRVs would be open. The -bypass keys are inserted, and switches 1S30 and 1S31 should be positioned to so that the no flow condition in and/or a high heat exchanger outlet temperature would not cause a reactor scram. The open TDHRVs in both branches would prevent running either TCS pump. When switches 1 S30 and 1 S31 are in relays are energized that close contacts that bypass the reactor safety system contacts actuated by -TCTE-2, and TCTE-3 in both legs of the TCS Protection System (See Figure 32). However, TDHR valve contacts in the TCS pump controllers would be open thereby preventing either TCS pump to run (See Figure 33). Page 61 of 190 ATTACHMENT 1 Figure 33 TDHRVs -TCS Pump Interlock Circuit Finally, if a single switch fails or is placed in the incorrect operating position, the worst condition that could be possible would be to have -TAs operating with one key switch in the 'Nonna!' position and the other key switch in In this condition, the high and low flow scrams and the high heat exchanger outlet temperature scram would be disabled for only one (1) of the two (2) reductant reactor safety system legs. Any other single switch mis-position or single switch failure with any operating condition would cause less protectlon to be bypassed. In addition, any single switch position or single switch failure with only one (1) TA operating would cause the operating TCS pump to stop due to the open TDHRVs in the secured TA cooling branch. The stopped TCS pump would reduce target flow and cause a reactor scram. Administratively, the keys to switches 1830 and 1831.will require Reactor Manager's permission to use. Since the MURR is not nonnally operated with bypass keys installed at the reactor console, the presence of keys will be a noticeable exception from nonnal operations. Therefore, operator attention to correct operation of the bypass keys will be peaked when they are implemented. 4.3.4 Target Cooling System Parameter Indication, Recording, and Alarm System The TCS Parameter Indication, Recording, and Alann System is designed to provide control room operators with indication of parameters convenient for operating the TCS, to record TCS parameter data for long-tenn retention or review, and to actuate alarms to alert the operator to abnormal parameters. No Page 62 of 190 ATTACHMENT 1 safety-related functions are associated with the TCS Parameter Indication, Recording, and Alann System. Every parameter is detected by a sensor. The sensor signal is converted to a 4-20mA current loop signal going to paperless chart recorders which also serve as parameter indication and alarm units. 4.3.5 Target Cooling System Secondary Coolant Control System The TCS Secondary Coolant Control System is designed to control the TCS secondary coolant circulation pump (SP-5) speed and the position of the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B). From the TCS Secondary Coolant Control System, the operator can adjust SP-5 to make course secondary coolant flow rate changes to the WCM. However, fine secondary coolant flow rate adjustment through a TCS heat exchanger is made by either S-3A or S-3B that are located in a secondary coolant pipe that bypasses the associated heat exchanger. For simplicity, further explanation of how the Target Cooling Automatic Temperature Control Valves (S-3A and S-3B) operate, S-3A will be assumed operating and its operation will be the example described. S-3B's controls and operation are the same as S-3A. During steady-state operation, SP-5 provides constant secondary coolant flow to the WCM. S-3A and its associated heat exchanger are in use, and S-3A is in 'Automatic' control. S-3A is located in a secondary coolant bypass line that bypasses the TCS heat exchanger. Therefore, as S-3A closes, less secondary coolant bypasses the heat exchanger resulting in more cooling flow to the heat exchanger. The opposite effect occurs as S-3A opens. The position of S-3A is controlled by a controller that is driven by the TCS heat exchanger outlet temperature transmitter TCTT-4. TCTT-4 receives its temperature measurement from resistance temperature detector TCTE-4. When slight changes to the TCS heat exchanger outlet temperature occur, S-3A moves to regulate the heat exchanger outlet temperature back to the desired set point. At the TCS Secondary Coolant Control System panel, the operator has manual switches and buttons to start, stop, and adjust the speed of SP-5. During normal operation, the operator will adjust SP-5 speed as necessary to ensure proper automatic operation of S-3A. S-3A has a 'Manual' control in which TCTT-4 does not affect the valve movement. In 'Manual', only the control room operator has buttons to open or close the valve. Manual control can be used for rapid changes in S-3A during operation, maintenance, or testing. 4.3.6 Target Cooling System Secondary Coolant Parameter Indication, Recording, and Alarm System The TCS Secondary Coolant Parameter Indication, Recording, and Alarm System is designed to provide control room operators with indication of parameters convenient for operating the TCS Secondary Coolant Control System, to record TCS secondary coolant parameter data for long-term retention or review, and to actuate alanns to alert the operator to abnormal parameters. No safety-related functions are associated with the TCS Secondary Coolant Parameter Indication, Recording, and Alarm System. Parameter current loop signals going to chart recorders which also serve as parameter indication and alarm units. Parameter inputs into the TCS Secondary Coolant Parameter Indication, Recording, and Page 63 of 190 ATTACHMENT 1 Alann System are multiple TCS Secondary Coolant Control System temperature, flow, and valve position signals. 4.3.7 N-16 Power Monitoring System The N-16 Power Monitoring System (PMS) is designed to provide control room operators with indication of reactor core power by means of measuring the amount of N-16 produced in the primary coolant system. Reactor fission power is directly proportional to the amount ofN-16 atoms produced by the fast neutron flux in the reactor. Therefore, reactor core power can be measured separate from the TEF power. The N-16 PMS has two (2) redundant detectors located in mechanical equipment room 114 and two (2) redundant displays located in the control room. The detectors will be placed in a shield with a collimator facing the primary coolant outlet pipe. The detectors are sealed stainless steel ionization chambers capable of measuring the dose rates at the designated measurement point. The N-16 PMS detectors are located on the reactor primary coolant hot leg piping at a distance corresponding to the time delay for primary coolant exiting the core to reach the measuring point of 1.35 times the N-16 half-life. This power measurement is calibrated by performing a series of power calorimetric calculations to determine reactor core power while not operating the TEF. This method of power measurement is used by several domestic and international research reactors and has been found to be stable and reliable. 4.3.8 Pool Coolant Monitoring System The Pool Coolant Monitoring System (PCMS) provides display and alarm functions both locally at the entryway to mechanical equipment room 114 and remotely in the control room based on I-131 activity concentration in the pool coolant system. Elevated 1-131 activity levels in the pool coolant system would be an early indication of a leaking target rod in the reactor pool. The PCMS has a detector in room 114 entryway. The PCMS continuously monitors pool coolant activity anytime the pool coolant system is in operation. The monitor receives 1 gpm of pool coolant from a Yi-inch vent line connection on the pool coolant circulation pump combined discharge header downstream of the pool coolant circulation pumps. The pool coolant water flows through a filter and a cation resin bed before it is monitored by a gamma scintillation detector in a 2.7-liter volume of the system. Once the pool coolant leaves the monitored volume, the coolant returns to the pool coolant system via Yi-inch vent piping on the pool water holdup tank. Given a 1 gpm flow rate, the overall time for the PCMS to detect 1-131 in the pool coolant system is approximately 13 minutes. The 13 minutes is based on the I-131 taking more than 11 minutes to get to the PCMS detector volume and almost two (2) minutes for the detector to recognize the increased I-131 concentration. Page 64 of 190 ATTACHMEN:T 1 5. Target Assembly Nuclear Design Analysis The target physics model predicts the production of Mo-99 as a function of the dimension, location and loading of the target rods and the neutron flux distribution in the target position. The physics model also provides the distribution of power densities in the. pellets for the thermal-hydraulic design of the TA. The target physics model incorporated both the T As and the MURR reactor core and examined the full range of MURR operations to develop the target physics design cases. 5.1 Analytical Methods The nuclear design calculations were conducted using Monte Carlo code MCNP6 (Reference 12) and ENDF/B-VII.1 (References 13 and 14) to obtain eigenvalue and neutron flux distributions. The MCNP6 calculations also produce heating rates for the reactor fuel elements and target rods, which are used for the nuclear design and analyses to obtain the TA power. For the determination of the TA uranium-235 enrichment distribution in MCNP6, 100 million source histories (1O,OOOx10,000 neutrons) were used to obtain converged eigenvalue and power distributions. 5.2 Target Assembly Physics Model The T As are loaded in graphite reflector positions . The physics design and analyses of the TAs utilized the "MURR 2015 reflector model" (Reference 15), which includes the reactor core and associated structures such as detailed modeling of the control blades (CB), regulating blade, beam tubes, beryllium (Be) reflector, graphite reflector, and experimental holes. The MCNP6 model of the MURR core is shown in Figure 34. The target rod physical dimensions used for the nuclear design calculations are summarized in Table 3. The physical dimensions utilized in the nuclear design are exactly the same as those of the mechanical design, which includes

target rods per target rod cartridge (flow channel), cartridge top and bottom fixture, cartridge guide rail, channel housing, and water plenum. The cartridge locking device, water diffuser, and incoming water pipe are not included in the physics model as they are expected to have little influence on the analyses. The physics model of the TA is shown in Figure 35 at the axial mid-plane, and target numbering is shown in Figure 36 for the baseline TA model. --The target rod model includes target pellets, pellet-clad gap, cladding, top plenum spring, and the top and bottom endcaps. In the physics model, -pellets are defined as a single axial node --The effective isotopic densities of. pellet consider -of open pore due to the pellet's design. Due to the complexity of the geometry of the top and bottom fixtures, the physics model of these components was designed to preserve the outer dimension and mass of the -material. Page 65of190 ATTACHMENT 1 Figure 34 MCNP6 Model of MURR with Driver Fuel and Reflector Element Numbers at Axial Mid-Plane Figure 35 MCNP6 Model of Target Assembly at Axial Mid-Plane Page 66 of 190 ATTACHMENT 1 Figure36 Target Rod Numbering for. Target Assemblies (baseline model) Figure 37 Model of the Target Cartridge Page 67 of 190 ATTACHMENT 1 5.2.1 MURR Driver Core States The physics design evaluates the target cartridge performance as well as the reactivity insertion to the MURR reactor core and driver fuel power changes due to TA loading and changes of the MURR core states. The TA design was conducted for the equilibrium core represented by its minimum, average and maximum bumup as summarized in Table 9. The extreme bumup core represents an anticipated driver fuel bumup distribution that can create the worst power peaking in both the driver fuel and target pellets, which has been generated from the maximum bumup core case without xenon. The critical CB position (distance from its fully inserted position) is the lowest at beginning-of-cycle (BOC) and withdraws as the driver fuel depletes and fission products accumulate in the core. For the typical 1-week operating cycle, the CB reaches its equilibrium height about 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months (Day 2 state) after startup and slowly withdraws until the end-of-cycle (EOC). Table 9 Definition of MURR Driver Core Burnup States (MWD) Driver Fuel Extreme Burnup Minimum Average Burnup Maximum Element Core Burnup Core Core Burnup Core <core ext> <core min> <core_avg> <core max> -Fl 0 0 19 3 F2 117 20 92 122 F3 67 18 60 68 F4 142 142 130 145 F5 0 0 19 3 F6 117 20 92 123 F7 67 18 60 68 F8 142 142 130 144 Core total (MWD) 652 360 600 676

Depending on reactor operating conditions, especially CB insertion height, the flux level in the reflector region could appreciably change. During 2014 -2015 operations, the lowest startup and highest shutdown critical position was recorded at CB heights of 32.69 cm ( 12.87 in) and 61.6 cm (24.25), respectively (Reference 16). These two CB positions are bounding values that include various reactor perturbations such as Be reflector installation, CB replacement and adjacent experiment loading/unloading. The CB movement during a 19-month period of operation in 2014 and 2015 is shown in Figure 38. The CB average age during this period is 4.7 to 5.7 years. In order to conservatively conduct the core and target power calculations, the limiting core configuration has been constructed to use a fresh Be reflector that will result in a higher neutron flux in the target region, in addition to the variations of driver fuel bumup. Page 68 of 190 ATTACHMENT 1

The CB age is tilted such th\lt A and D are fresh and B and C are eight (8) years old, which will result in a higher power peaking in the TA.

The CB tip position is tilted so that the tip position of CB B and C is higher than that of A and D by 2.54 cm (1 in), which results in a higher power peaking in the TA.

The central flux trap is loaded with sample materials.

Equilibrium xenon is used for the maximum and average bumup driver fuels.

No xenon is used for the extreme and minimum bumup driver fuels. Figure 38 MURR Control Blade Travel between January 13, 2014 and September 15, 2015 Typically, the CB travels over 10 cm (3.94 in) during the first two (2) days after startup and then slowly continues to withdraw. The driver fuel elements also have two (2) distinct states: clean fuel and equilibrium xenon. The xenon-free core bumup is provided in Table 10. The core burriU:p of an equilibrium xenon state core is approximately 20 MWD higher than that of the xenon-free core. Because the pellet enrichment is fixed at nominally-the target cartridge design process searched for the target rod position that satisfies the thermal design limits of the target rod and cartridge during normal operating conditions (accident conditions presume operation at the allowable design limits). The selected target rod position is from the core center. Page 69 of 190 ATTACHMENT 1 The critical CB positions and expected CB travel range of the equilibrium core are summarized in Table 10 and Table 11. When the TAs are loaded, the CBs are inserted deeper into the core. The estimated additional CB insertion depth is 3 to 9 cm (1.2 to 3.5 in) depending on the core states. When these offset values are applied to the lowest and highest CB position during 2014 to. 2015 operation, the lowest and highest critical CB position with a two (2) TA loading is expected to be 33.7 cm (13.27 in) and 51.9 cm (20.43 in), respectively (Table 11). Here, the CB offset values of <core_ext> and <core_max> were applied to BOC and EOC cases, respectively, while that of <core_avg> was approximately used for Day-2 CB position. Table 10 Critical Control Blade Positions ' ' -"" ' .: Extreriie . Ma.ximriin *

l\firiimurli .. ... ' . . .. . . * . .*. * * *' eore ' '-'" ". '\,-* Xenon state no xenon no xenon equilibrium equilibrium Regulating blade (cm) 25.4 25.4 38.1 38.1 Critical CB position without 41.67 35.09 59.87 65.00 target assembly Critical CB position with full (2 38.75 32.75 52.89 55.28 fresh) target assemblies loaded Control blade offset (cm) 2.92 2.34 6.98 9.72 Table 11 Expected Control Blade Travelling Range for Target-Loaded Core * . Critical c;.B * * .. Expected Critical' CB Range with Target Loading

Target Loading

Lowest *. Highest . Lo.west Highest BOC 36.6 42.6 33.7 39.6 Day-2 49.l 58.8 42.1 51.8 EOC 53.6 61.6 43.9 51.9 5.2.2 TargetAssembly Criticality A conservative estimate of Keff for -T As was perfomied using the MURR core model by replacing the driver fuel with water and assuming the control and regulating blades are completely withdrawn from The calculated values of Keff for the target cartridge located in reflector position , indicating that the T As will be subcritical with a large margin for uncertainty. The uncertainties are given in 2 standard deviations (2a) as a 95% confidence level. Page 70of190 ATTACHMENT 1 5.2.3 Impact of Target Assembly Loading on MURR Reactor Core Loading

T As in the MURR graphite reflector region causes perturbations in neutronic and thermal performance of the MURR core as follows:

The MURR core excess reactivity will increase slightly, which will be compensated by the reactivity control devices.

The neutron flux and power of the adjacent. fuel element will increase even though the total core power is maintained at 10 MW.

The thermal power generated by the target rods will not affect the reactor core primary or pool cooling capability. The heat generated by the targets that is discharged to the reactor pool will be adequately removed by the reactor pool coolant system without impacting the pool temperature. 5.2.4 Reactivity Insertion The reactivity insertion due to target cartridges being loaded in graphite reflector positions --is summarized in Table 12. The reactivity worth was calculated by replacing the water in the TA cartridge with fresh, clean, cold target rods. Under the hot operating condition, the reactivity insertion due to a single cartridge loading is less than -* The maximum reactivity insertion due to a hot target rod is the same as that of the cold target rod for all core burnup states. Table 12 K.,ff and Reactivity Insertion Values for Single Target Assembly filled -** -* with structure and water K.,ff (%) Ke ff (%) (K.rr) Maximum with equilibrium 0.99368 -* -* xenon Average with equilibrium 0.99392 -* -* xenon Minimum without xenon 0.99385 -* -* Extreme without xenon 0.99382 -* -* 5.2.5 Driver Fuel Power Peaking The target cartridge loading in the graphite reflector positions has a relatively small impact on power peaking of the MURR core driver fuel. Table 13 compares the distribution of MURR driver fuel power with and without TA loading and provides the location of the largest peaking factors in the reactor core. For most of core states, the fuel element * (closest to the TAs, see Figure 34 for driver fuel identification) peaking factor is the highest because the fuel burnup is the lowest and the fuel is located close to the TA. For the F5 fuel element, the peaking factor of the inner plate is always higher than that Page 71 of 190 ATTACHMENT 1 of the outer plate. When -T As are loaded, the increase of peaking factor is less than 4% for the inner plate, while the maximum increase of peaking factor is 11 % for the outer plate. However, the absolute value of the outer plate peaking factor is always less than that of the inner plate peaking factor. Table 13 MURR Core Maximum Power Peaking Due to Target Assembly Loading Inner Fuel Plate Outer Fuel Plate Number of Target Rods Axial Node1 Peaking Axial Node Peaking (Fuel Element) Factor * (Fuel Element) Factor Maximum with I 13 (F5) 2.541 13 (Fl) 2.064 equilibrium

11 (F5) 2.637 11 (F5) 2.218 xenon Average with I 11 (F5) 2.426 12 (Fl) 1.999 equilibrium

8 (F5) 2.511 11 (F5) xenon 2.155 Minimum I 10 (Fl) 2.968 9 (F6) 2.444 without xenon

9 (F5) 3.031 9 (F6) 2.565 Extreme without I 10 (F5) 2.905 9 (Fl) 2.200 xenon

10 (F5) 2.976 8 (F5) 2.397 1 Axial nodes I and 24 correspond to the bottom and top of MURR driver fuel plate, equally spaced. 5.2.6 Reactivity Coefficients Though the core power distribution and power peaking factors are altered when two (2) target cartridges are loaded in the graphite reflector region, the core reactivity characteristics do not change significantly due to target rods being loaded in the reflector region. The reactivity coefficients of the core with -fresh T As were calculated for the fuel temperature coefficient, coolant temperature coefficient and void reactivity and are summarized in Table 14. The driver fuel temperature coefficient was calculated by increasing the fuel temperature by 906.4 °C (1663.5 °F), i.e. from 20.44 °C (68.8 °F) to 926.84 °C (1700.3 °F). The fuel temperature coefficients are consistent with core bumup state and kept negative. Due to the limitations in cross section data, however, the temperature dependence of lumped fission products (all fission products except 1351, 135Xe, 149Pm, 149Sm) was not considered. The coolant temperature coefficient was calculated by changing the coolant temperature from 20.44 °C (68.8 °F) to 76.84 °C (170.3 °F), where the coolant density changes from 0.99815 to 0.97378 g/cm3. The principal cross section data used for the calculations are basically the same for both the lower and higher coolant temperature conditions, but the S( a,p) thermal scattering was correctly treated. The coolant void reactivity of the core was calculated by removing 99.9% of coolant from the core. Page 72 of 190 ATTACHMENT 1 The statistical uncertainty (20) of the reactivity coefficient is when the error propagation rule is used. The least negative values are -1.03 x 1 o-6 and -1.69 x 10*4 Ak/k/°C for the fuel and coolant temperature coefficient, respectively. Negative reactivity coefficients of coolant void and temperature limit the peak power during a reactivity transient. Using the lowest magnitude of these negative reactivity coefficients for a positive reactivity insertion accident analysis is conservative because the subsequent power transient is larger than if nominal values were used. Table 14 Reactivity Coefficients of the Reactor Core with

Fresh Target Assemblies CoreBurnup Fuel Temperature Coolant Temperature Coolant Void Coefficient Coefficient Reaetivity State (Ak/k/oC) (Ak/k/0C) (Ak/k/o/ovoid) Maximum with -1.20 x 10"6 -1.69 x 10*4 -3.40 x 10*3 equilibrium xenon Average with -1.16 x 10"6 -1.73 x 10*4 -3.40 x 10*3 equilibrium xenon Minimum without -1.27 x 10"6 -1.85 x 10*4 -3.40 x 10*3 xenon Extreme without -1.03 x 10"6 -1.79 x 10*4 -3.40 x 10*3 xenon 5.2.7 Reactivity Device Worth The impact of TA loading on the existing reactivity control devices (control and regulating blades) were estimated for their reactivity worth and subcriticality margin as summarized in Table 15. Reactivity worth of the regulating blade was calculated from fully inserted and fully withdrawn conditions while the CB is kept at its critical position. For the selected core states, the lowest worth of the regulating blade is 3.41 x 10*3 Ak/k. The CB subcriticality margin was at first calculated for the average bumup core. Among four (4) CB's (A, B, C and D), the subcriticality margin of C is the lowest even though it is almost the same as that of B. The subcriticality margin increases as the core bumup increases due to effective neutron flux changes. The lowest subcriticality margin with the most reactive CB (i.e. A) and regulating blade fully withdrawn is 0.055 Ak/k for the minimum bumup core. Page 73 of 190 ATTACHMENT 1 Table 15 Reactivity Control Device Worth with

Fresh Target Assemblies *'" ' *,' S:iJ.bcfiticalitY Margin , .. .. . *,' Blade Wor,tli .. of C.oritiol * ., (Aklk,) " {Ak:/k). .. Max.imum with equilibrium xenon 0.109 3.55 x 10-3 Average with equilibrium xenon 0.106 3.41 x 10-3 Minimum without xenon 0.055 3.69 x 10-3 Extreme without xenon 0.076 4.02 x 10-3 5.2.8 Core Excess Reactivity The excess reactivity of the core with -fresh TAs was calculated for the minimum bumup core which has the largest excess reactivity. All CB' s and regulating blade are fully withdrawn from the core. The excess reactivity of the cold core (all at room temperature) is 0.072 Ak/k, while that of the hot operating core is 0.067 Ak/k. 5.2.9 Kinetic Parameters The kinetic behavior of the core is dominated by the driver fuel elements, which is slightly perturbed by the presence of-T As in the graphite reflector region. The effective (or adjoint weighted) neutron generation time (Aetr) and effective delayed neutron fraction were generated for different bumup states of the core with and without (Reference 17) T As. as summarized in Table 16. The neutron generation time tends to increases when the -T As are loaded, but very slightly. The effective delayed neutron fractions of the core with and without T As are very close to each other within the uncertainty range (+/-2cr). From the viewpoint of point kinetics, the TA loading won't deteriorate the slope of power increase (or inverse reactivity period) during the reactivity induced transient. Page 74of190 ATTACHMENT 1 Table 16 Kinetic Parameters of the Reactor Core with and without Target Assemblies Targ.et CoreBurnup *A.rr STD ... STD Loading State (µsec) * (fo) ' *Perr * (fo) Maximum 62.3 0.273 0.00723 0.00015 No Target Average 60.6 0.278 0.00731 0.00015 Assemblies Minimum 53.7 0.244 0.00749 0.00016 Extreme 57.0 0.248 0.00732 0.00015 Maximum 62.7 0.237 0.00730 0.00013 .Target Average 61.5 0.235 0.00745 0.00013 Assemblies Minimum 54.7 0.215 0.00766 0.00014 Extreme 58.4 0.220 0.00769 0.00014 5.2.10 Structural Component Heating The structural component heating was estimated by MCNP6 code using the kinetic energy deposited by the fission fragments, prompt neutrons and delayed neutrons. In addition to the neutron data used for neutron flux calculation, the heating calculation uses photon-atomic data (mcplib04) (Reference 18), photon-nuclear data (endf7u) (Reference 19), electron data (el03) (Reference 20), and delayed neutron/gamma data (cinder.dat, delay_library.dat, cindergl.dat, delay_library_v2.dat) (References 21 and 22). MCNP6 adopts the ClNDER'90 model to calculate delayed particle emissions from all of the radionuclides in the decay chain for a fission product. The photon emission time for delayed-neutron and delayed-photon emission was adjusted to 105 sec (the default value is 1010 sec), under which the statistical uncertainty (lcr) of the calculated heat deposited in the Be reflector is -5%. The Be reflector heating is higher for the maximum burnup core by 3.3% when compared with the minimum bumup core. For the maximum burnup core, the Be reflector heating is -with and without two (2) TAs, respectively. Page 75of190 ATTACHMENT 1 Table 17 Component Radiation Heating Due to Target Assembly Loading for Maximum Burnup Core Target Neutron Photon . Electron Heating Total Component Heating Heating Loading (kW) (kW) (kW) (kW) No Target Be reflector * * * .. Assemblies Be reflector * * * .. .Target -* * *

Assemblies -* * *

5.3 Target Assembly Flux and Power 5.3.1 Fresh Target Assemblies The neutron flux in a TA is dominated by incoming neutrons from the reactor core even though the TA produces neutrons from fission reactions. Figure 39 and Figure 40 show 4-group axial neutron flux distributions of the target rods -i.e. the middle of each

target rod bundle (see Figure 36 for target rod numbering), when two fresh target cartridges are loaded in graphite reflector positions -. The upper energy boundaries of the 4-group are 20 MeV (Group 1), 0.1 MeV (Group 2), 5.53 keV (Group 3), and 0.625 eV (Group 4). The solid and hollow symbols in each figure indicate the MURR core states, i.e. <core_ext> and <core_max>, respectively. Axial nodes 1 and 25 correspond to the bottom and top pellet, respectively. As can be seen, the neutron flux drops at the bottom and top section of the target rod. The neutron flux is suppressed even more in the top section due to the CBs, which is relaxed to a certain extent when the CBs are pulled out in the maximum bumup core. Figure 41 shows neutron flux in the azimuthal direction, from target rod at axial middle plane. Page 76of190 ATTACHMENT 1

Group 1

Group 2 * *

I a a

Group 3

a * *

a *

Group 4 a a

a a *

a

core_max> a a a * *

a a *

9 * * * *

0 &

t

0 *

0

0

i t $ i i *

0 5 10 15 20 25 Axial node Figure 39 Target Rod I Axial Neutron Flux Distribution

Group 1

Group 2 a

Group 3 a

a

Group 4 a a 'iii' a N a

a E <core max

c -* a )( ::I

a i;:::: c a * -* ::I * * * * *

Q)

0 * *

z * * *

0 t &

a 0 * *

0 *

I t

i i *

5 10 15 20 25 Axial node Figure 40 Target Rod. Axial Neutron Flux Distribution Page 77 of 190 ATTACHMENT 1 <core_ext> * * * * * *

I! * *

D D D D D B B D D D D

D

D I D <core_max> B

D I t e e t s I I I

I I I I I * * * * * * *

8 * * * * * * * * *

Rod number Figure 41 Target Assembly Azimuthal Neutron Flux Distribution

D I *

D

a *

D

8

Calculated target power and pellet linear power when -fresh T As are being irradiated are given in Table 18. Figure 42 and Figure 43 show the pellet linear power envelope of the extreme and maximum burnup cores respectively. The axial power profile is initially bottom-peaked and gradually changes to middle-peaked shape as the regulating and control blades are withdrawn from the core. The distribution of pellet linear power is shown in Figure 44 for the four (4) different MURR core states. For the most probable operating core condition, i.e. the average burnup core, the peak pellet linear power and total TA power are , respectively. Table 18 Calculated Target Power Level and Linear Power Peak Linear Power (kW/m) Target Rod Number Axial Node* Extreme Burnup Core I Minimum Burnup Core I *Axial nodes I and 25 correspond to bottom and top node, equally spaced. Page 78 of 190 Average Burnup Core I Maximum Burnup Core I ATTACHMENT 1 0 5 10 15 20 25 Axial node Figure 42 Power Envelope of the Base Target Loading for the Extreme Burnup Core Case 0 5 10 15 20 25 Axial node Figure 43 Power Envelope of the Base Target Loading for the Maximum Burn up Core Case Page 79 of 190

.------------------------------------------------------------------G> c. 'O .... Cl> .c E :::i z * <core_max> * <core_avg> * <core_min> * <core_ext> *

ATTACHMENT 1 * * * * * * * * ** ****** Pellet linear power (kW/m) Figure 44 Target Assembly Pellet Linear Power Distribution 5.3.2 Sensitivity to Control Blade Position The peak linear power of the target rod is dominated by the CB insertion depth. The sensitivity of peak linear power to CB position was estimated for the maximum, average, minimum and extreme core bumup states. The CB position represents the core response to the various reactivity perturbations seen by the reactor core, including fuel depletion and material aging. Sensitivity calculations have been conducted for a wide range of CB positions beyond the expected minimum and maximum critical position for the target-loaded core. In this simulation, the regulating blade was fixed at its typical position: 25.4 cm (10 in) for the extreme and minimum bumup core and 38.1 cm (15 in) for the average and maximum bumup core. It should be noted that the CB movement is synchronized with the regulating blade in real operation so that the actual operating range of CB will be narrower than the simulated one. The results are summarized in Table 19 for the total target power, peak target linear power and driver fuel peaking factors. The total target power is relatively low for the BOC condition, which is a relatively short tenn period. For the equilibrium fuel conditions (average and maximum bumup cores), the target power stays below -The peak linear power of the target rod is maintained between for the simulated equilibrium core states and CB position. The variations in total target power and peak linear power versus CB position are shown in Figure 45 and Figure 46, respectively. The MURR driver fuel element power peaking occurs in fuel element *. The driver fuel element peaking is higher for the BOC state when the CB and regulating blade are deep into the core. The calculated maximum peaking factor is 3.231 for the inner plate. For the outer plate that faces the target, Page 80 of 190 ATTACHMENT 1 the maximum peaking factor is 2.636, which is lower than that of the inner plate for all core state and CB positions. The variations of driver fuel element peaking factor vs CB position are shown in Figure 47. The effect of CB position has also been studied for the critical core configurations by changing the CB average age from 0 to 8 years. For all the critical core cases, the calculated total target power and the maximum linear power are less than those of non-critical core cases. For the extreme burnup core, the maximum linear power is for the non-critical and critical core, respectively, when the average CB age is the same ( 4 years) for both cases. The extreme burnup core becomes critical at CB height of 32.28 cm (12.71 in) when the age of all the CBs is 8 years .. The maximum linear power of the extreme burnup core with CB at cm (12.6 in) is for the non-critical and critical core, respectively. It should be noted that the CB age tilt is 8 years for the non-critical case while it is 0 for the critical case, which reduces the neutron population in the TA to a certain extent; and both the total target power and the maximum linear power are reduced in the critical core when compared with the critical core. The critical core calculations have shown that the non-critical core calculations conservatively estimate the target power and the maximum linear power. However, it should also be noted that the critical core simulation here is limited by the range of CB age and its tilt and therefore, the variation of CB position is smaller when compared with the expected CB traveling range. This means that variation of CB age (as well as* CB age tilt) is not sufficient to realistically model all the critical core cases. It is also true that in the actual core the CB position could be even higher or lower than those used for the critical core calculations due to Be reflector aging and other reactivity perturbations. Finally, it is also reasonable to assume.that the target power and its axial shape are dominated by CB position. Therefore, the limiting core configuration has been selected from the non-critical core configurations. The maximum linear power found from the sensitivity calculation is -for the extreme burnup core with the CBs positioned at 30 cm (11.81 in) and the regulating blade positioned at 25.4 cm (10 in), which is used for the target thermal-hydraulics and CHF analyses. The maximum TA power is -for the maximum burnup core with CB at 44 cm (17.32 in) and regulating blade at 38.1 cm (15 in), which is the recommended case for the TA cooling analysis. Page 81of190 ATTACHMENT 1 Table 19 Target Assembly Power and Core Peaking Factors CB Position Target Power Peak Linear Inner Plat.e Outer Plate Core State Power (cm) * .(kW) (kW/m) Peaking Peaking * .. * .. .. * .. * .. .. * .. * .. .. Maximum * .. * .. .. bumup with * .. * .. .. equilibrium * .. * .. .. xenon * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. Average * .. * .. .. bumup with * .. * .. .. equilibrium * .. * .. .. xenon * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. Minimum * .. * .. .. burn up * .. * .. .. without xenon * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. * .. * .. .. Extreme * .. * .. .. burn up * .. * .. .. without xenon * .. * .. .. * .. * .. .. * .. * .. .. Page 82of190

'E i .¥ -... Q) a. ... ns Q) c .¥ ns Q) Q. 30 ATTACHMENT 1 Equilibrium xenon case a a 0 0 a a 0

a 0 0 *

a 0

a * )( )( )( No xenon case 35 40 45 50 55 60 CB position (cm) Figure 45 Variation of Target Power vs. Control Blade Position 0 )( 0 )( 0 )( 0 )( No xenon case 35 a )( 40 Equilibrium xenon case 0 a 45 ' 8 9 50 CB position (cm) Figure 46 55

a 0 60 Variation of Peak Linear Power vs. Control Blade Position Page 83 of 190 0 65

65 ATTACHMENT 1 Inner plate 0 -CJ (Equilibrium xenon case) CV -0 C) If 0 0 c a a 0 0 0 a 0 0 .II:

CV I * *

CD * * *

c.. * * * * *

CD -CV c.. Qi ::I -CD > *;:: c 45 50 55 60 65 CB position (cm) Figure 47 Variation of Driver Fuel Element Peaking Factor vs. Control Blade Position The effect of regulating blade movement was assessed for the critical core. Two (2) regulating blade positions, i.e. low (25.4 cm) and high (38.1 cm), were considered while the CBs were at their critical positions. The results in Table 20 show that the pellet peak linear power and total TA power are . The maximum inner and outer plate peaking factors of the driver fuel are 3.043 and 2.576, respectively, for the minimum bumup core. Page 84 of 190 ATTACHMENT 1 Table 20 Effect of Regulating Blade on Pellet Peak Linear Power and Target Assembly Power Regulating Target Peak Linear Inner Plate Outer Plate. Core State Blade

Power Power . (cm) -(kW) (kW/m) Peaking .. Peaking Maximum bumup * ..

2.637 2.218 with equilibrium .. ..

2.641 2.147 xenon * ..

2.638 2.046 Average bumup * ..

2.511 2.155 with equilibrium .. ..

2.542 2.093 xenon * ..

2.528 1.973 * ..

3.037 2.576 Minimum bumup .. ..

3.043 2.563 without xenon * ..

3.031 2.565 * ..

2.988 2.531 Extreme bumup .. ..

2.999 2.507 without xenon ** ..

2.976 2.397 5.3.3 Uncertainties The neutronics analysis has an inherent uncertainty due to the solution method employed in the MCNP6 code. The standard deviation (1 cr) of the pellet power is -for almost all the pellet numerical nodes when 100 million particles are used for the calculation. For the total target power, the standard deviation (lcr) is as small as -* Other uncertainties considered are impurities of constituent materials, pellet density, fissile content and target rod position due to manufacturing tolerance. Table 21 shows impurities of of the target pellet, cladding, cartridge and neutron shield. For the and -, the impurity was assumed to be Table 21 Material Impurities for Uncertainty Analysis Uranium (wt%).* 232u 2 x 10-7 NIA NIA NIA 234u 0.26 NIA NIA NIA 236u 0.46 NIA NIA NIA Boron 3ppm 0.00005 10 10 Page 85 of 190 ATTACHMENT 1 The uncertainties of the peak linear power and total target power were estimated for the four (4) critical cores. In order to estimate the sensitivity of the core and target performance parameters, the uncertainties of design parameters were defined as follows:

For the statistical uncertainty of the solution method, +/-2cr value is used to estimate the eigenvalue, total target power and peak linear power with a 95% confidence level.

The impurity always reduces the target power and is regarded as a bias. The uncertainty is estimated as 95% of the performance parameter change due to the maximum impurity level of the target uranium, assuming a flat distribution of impurity between its minimum and maximum values.

The manufacturing tolerances of the pellet density and fissile content are respetively. The target performance is relatively insensitive to the variation of these parameters. For conservatism, the tolerance value is taken as the uncertainty. In this simulation, only a positive value is used as a bias to estimate the power increases.

The uncertainty of the target rod position is , i.e. approximately a 2cr value when a triangular distribution is assumed between the minimum and maximum values with the mode (average) at 0. The. cm means that the rod is closer to the core by

cm. Though the target rod could be either closer to or farther from the core, -is used as a bias to estimate the power increase of the target. The estimated uncertainties (positive components) of the core eigenvalue, peak linear power and total target power due to the statistical and manufacturing uncertainties are summarized in Table 22, Table 23, and Table 24, respectively. The impact of the pellet fabrication density and fissile content uncertainty on the core eigenvalue .and target peak linear power is very small, which makes it difficult to obtain consistent results by direct perturbation calculations. Therefore, large manufacturing uncertainties

were used for the direct calculation and the results were linearly interpolated. The uncertainty (positive variation) of the eigenvalue due to manufacturing tolerances (pellet density and fissile content) is estimated to be less than . The impact of target rod position uncertainty mostly dominates the total uncertainty for all core states. For the peak linear power, the statistical uncertainty prevails over of the pellet density and fissile content, while it is comparable to the target rod position uncertainty. Arithmetic summation of these uncertainties, excluding the impurity effect, results in a total uncertainty of-for the peak linear power in the extreme bumup core, which is very conservative. Specifically, for the extreme burnup core, the uncertainty due to the most dominant contributor, i.e. rod position uncertainty, is -For the total target power, the statistical uncertainty is relatively small, while the target rod position uncertainty dominates the total uncertainty. The estimated maximum increase of target power due to uncertainties associated with manufacturing tolerances (pellet density and enrichment) is Page 86of190 ATTACHMENT 1

The total uncertainties of the key performance parameters were estimated as a product of statistical error and root-mean-square (RMS) of uncertainties due to fabrication density, enrichment, and target rod position. . Here, the uncertainty due to impurity is neglected for conservatism. For the limiting core configurations, the total uncertainty (excluding the impurity effect) is applied to the target design power such as: where P and Po are power with and without uncertainty, respectively. Subscripts s, d, e, and p refer to the uncertainties due to statistical, density, enrichment, and position, respectively.

For the eigenvalue, the estimated uncertainty is +/-0.00048 for the maximum burnup core.

For the total target power, the estimated uncertainty is -for the maximum burnup core. The estimated upper bound for the total target fission power is then -when considering uncertainties (cf. -without uncertainty).

For the peak linear power, the estimated uncertainty is -for the extreme burnup core. The estimated upper bound for the peak linear power is then -when considering uncertainties (cf. -without uncertainty). Table 22 Uncertainties in Core Eigenvalue (Bias) CoreBurnup Statistical . Impurity Fabrication Fissile Target Rod Uncertainty Density Content Position State (pcm) (pcm) (pcm) (pcin) (pcni) Maximum with * *

I

equilibrium xenon Average with * *

I

equilibrium xenon Minimum without

I I I

xenon Extreme without

I I I

xenon Page 87of190 ATTACHMENT 1 Table 23 Uncertainties in Peak Linear Power (Bias) CoreBurnup Statistical Impurity Fabrication Fissile Content Target Rod Uncertainty Density Position State (%) (%) . (%) (%) (%) Maximum with * -* *

equilibrium xenon Average with * -* *

equilibrium xenon Minimum wi(hout * -* *

xenon Extreme without * -* *

xenon Table 24 Uncertainties in Target Power (Bias) CoreBurnup Statistical Impurity Fabrication Fissile Target Rod Uncertainty Density Content Position State (%) (%) (%) (%) (%) Maximum with * -* *

equilibrium xenon Average with * -* *

equilibrium xenon Minimum without * -* *

xenon Extreme without * -* *

xenon 5.3.4 Nominal Operating Cycle (Staggered Loading Pattern) The nominal operating cycle will load and remove

target rods per week for processing. In a nominal operating cycle, Page 88 of 190 ATTACHMENT 1 Figure 48 and Figure 49 show 4-group axial neutron flux distributions of the target rods 6 and 17, respectively, when a fresh TA is loaded in reflector position -of the reference and extreme bumup core. The axial flux shape of the staggered target loading is almost the same as that of the fresh target loading. Figure 50 shows the azimuthal flux variation on the axial middle plane, which shows a slight decrease of neutron flux level in position -'ii) C'li E c -)( ::I i;:::: c 0 ... .... ::I Q) z *

II Iii <core_ext> *

a a

a

a

a <core_max> a

a * * * *

t 9 * * * * *

I *

0

0 I

I I 0 5 10 Axial node Figure 48 -Axial Neutron Flux For the nominal loading with a fresh TA

Group 1

Group 2 a

Group 3 a

a

Group 4 a

a

a *

a

0

0 * *

0 A 0 : i ' ' 15 20 25

the 4-group axial neutron flux distributions of target rods -are shown in Figure 51 and Figure 52, respectively. The azimuthal fluxes are shown in Figure 53. Page 89 of 190 ATTACHMENT 1 * * *

a

Group 1 *

Group 2

a a a I a a a

a

Group 3 a

Group 4 a

a

a a <core_max> a a a

a I I a * * *

a I a * * * . .. g *

0 *

I 0 ..

0

0 .. 0

0 :

0 0 i 8 *

0 0 *

5 10 15 20 25 Axial node Figure 49 * * <core_ext> *

II

a a a a * * * * * * *

a a *

a a a a a

a a a II a <core_max> a *

a

a a

Group 1

Group 2

Group 3

Group 4 ' A *

A 6 I I I I I I I I I 9 t e

0 0 I I 0 * * * * * * * * * * * * * * * * * * * *

Rod number Figure 50 Page 90 of 190 ATTACHMENT 1

Group 1 *

l!I 8 a *Group 2 a a a

Group 3

a *Group 4 'iii'

a

a N

E

a <core_max> a .!:? a a .s )( a

a :J a c 0 * ... -* I * * :J

Cl) z * * * .. 0 * * & 0 0 *

t ..

0 *

0

0 i 0

0 0 0 * *

0 *

5 10 15 20 25 Axial node Figure 51 -Axial Neutron Flux

Group 1

Group 2

II *

a a

Group 3 a

a

Group 4 'iii'

a N a

E * !:!::core max a .!:?

a .s a

a )( * :J a

c a E * -* :J * *

i : t t

Cl)

z * *

0 A : . 0 0 t .. 0 *

0 t

i 0 0

0 i *

i 5 10 15 20 25 Axial node Figure 52 Page 91of190 ATTACHMENT 1 Axial node 12 <core_ext> * ! * * * *

I D D * * * *

D D Iii

D D D D *

D <core_max> D *

D D D

l'i D D

D D E II D £:!

Group 1 .:. )( *Group 2 :I

Group 3 II= c: ,. Group 4 0 I

I

8 t t t

t GI * * *

I I I I I I z I I 0 0 0

0 * * * * * * * * * * * * * * * * * * * *

2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Rod number Figure 53 TA power and rod linear power of the staggered TA loading is summarized in Table 25 when the fresh assembly is in graphite reflector position -* The pellet linear power envelopes -are shown in Figure 54 and Figure 55 for the extreme and maximum bumup core, respectively. The distribution of pellet linear power is shown in Figure 56. The axial power profile -is the same as that of the fresh target loading, but the overall linear power is slightly reduced in the burned assembly. For the most probable operating condition, i.e. average bumup core, the power of the TA Page 92 of 190 ATTACHMENT 1 Table 2S Calculated Target Power Level and Pellet Linear Power Core Burnup Target Power Staggered Loading Staggered Loading State (fresh target in SA) (fresh target in SB) Assembly Power (kW) .. .. Maximum with equilibrium Peak Linear Power (kW/m) xenon Target Rod Number I

Axial Node Assembly Power (kW) .. .. Average with equilibrium Peak Linear Power (kW /m) xenon Target Rod Number I

Axial Node Assembly Power (kW) .. .. Minimum without xenon Peak Linear Power (kW/m) Target Rod Number I

Axial Node Assembly Power (kW) .. Extreme without xenon Peak Linear Power (kW/m) Target Rod Number I

Axial Node Page 93 of 190 0 Power Envelope E c. .. "' CD c ::; Power Envelope 5 ATTACHMENT 1 10 15 20 25 Axial node Figure 54 Axial node Figure 55 Page 94 of 190 ATTACHMENT 1 * <core_max> * <core_avg> * <core_min> * <core_ext> * * * ** * ** * *** ** * * * * . .. .. * * * ** Figure 56 Target Assembly Pellet Linear Power Distribution ** * * ** * * * ** * * ** * ** * * *

The pellet linear power envelopes of the staggered loading with a fresh TA are shown in Figure 57 and Figure 58 for the extreme and maximum burnup core, respectively, and their distribution is plotted in Figure 59. Page 95 of 190 E' i .II: -... ; 0 a. ... nl m c :J ATTACHMENT 1 Axial node Figure 57 Axial node Figure 58 Page 96 of 190 x x 25 x 25 ATTACHMENT 1 * <core_max> * <core_avg> * <core_min> J!l * <core_ext> * ..! ** * *

Q) c.. -0 ...

Q) ..c E ::J z

Pellet linear power {kW/m) Figure 59 Target Assembly Pellet Linear Power Distribution 5.3.5 Target Assembly Flux and Power of Partial Target Assembly Loading The number of target rods loaded in a TA may be less than

during commissioning. For that purpose, the neutron flux and power distribution of the TA were simulated for the case of a Other rods positions are loaded with filler rods to avoid excessive neutron thennalization while maintaining the nominal flow conditions. The 4-group axial neutron flux distributions of target rods -are shown in Figure 60 and Figure 61, respectively. The azimuthal fluxes are shown in Figure 62. TA power and pellet linear power are summarized below in Table 26. The pellet linear power envelopes of the extreme and maximum bumup core are shown in Figure 63 and Figure 64, and their distribution is plotted in Figure 65. Page 97 of 190 ATTACHMENT 1 Table 26 Calculated Target Power Level and Pellet Linear Power for a Reference ITarget Rod Partial Loading Extreme Minimum Average Maximum Burnup Core Burnup Core Burnup Core Burnup Core Peak Linear Power (kW/m) Rod Number Axial Node <core_ext> * *

a a c 2 :; Cl> z Cl 5 * * * * * *

a -=-+a a -a--* 0 a a a a <core_max> 3 3 *

10 15 Axial node Figure 60

Group 1 a *Group 2

Group 3

a *Group 4

a

a

a a

A A A A * ... A *

i i ... A * * *

20 -Axial Neutron Flux for Partial Loading Case Page 98 of 190 25 ATTACHMENT 1

Group 1 * * * . a a

Group 2 * *

a

Group 3 a

a 0 a

Group 4 iii' a a I

N

a E o-a.....E 11 <core_max> I

u a -.s a

a )(

a ::i a

c a

a E * *

t ' t

t A A * .... .4 ::i i A

A Cl> A * * *

e

A z

0 0 0 *

i

A 0 0 i 0

g e * $ * * * * * * * * * * * $

i $ $ ---5 10 15 20 25 Axial node Figure 61 Axial Neutron Flux for Partial Loading Case Axial node 12 <core_ext> * * *

a *

a a a a a ... ... ... -;-... ... ... <core_max> ... A A N A A A ... * ... A A E A A t

t A A .:.. ... ... )( A A ::I c ... t e A ! t t t '$ I

8 8 G> 9 I * *

g * *

g 8 8 z

0 0 0 i *

9 8 0 0 II a II 0 0 *

I II a a II * *

3 * * * * * * * * * * * * * * *

Figure 62 Target Assembly Azimuthal Neutron Flux for Partial Loading Case Page 99 of 190 ATTACHMENT 1 10 '

Axial node Figure 63 15 J * * *

I I 20 25 Power Envelope for Partial Loading for the Extreme Burnup Core Case i *

i I

5 10 15 20 25 Axial node Figure 64 Power Envelope of the Partial Loading for the Maximum Burn up Core Case Page 100of190 T * <core_avg> * <core_min> A <core_ext> ATTACHMENT 1 AA

Pellet linear power (kW/m) Figure 65 Target Assembly Pellet Linear Power Distribution for Partial Loading Case 5.3.6 Target Assembly Material Depletion As the fissile material bums in the target rods from irradiation by the incident neutron flux from the reactor core, the TA power steadily decreases. Table 27 lists the isotopic mass of -T As for the average bumup core, which is the most likely core state. As the fissile uranium depletes, the TA power also decreases linearly as shown in Figure 66. At the end of a 2-week irradiation, the TA power drops to -' which corresponds to a -reduction when compared to the initial power of -for the commercial loading. The target pellet bumup is Table 27 Material Content and Burnup of the Base Target Loading for Average Burnup Core State Calendar Operation Burnup 23su 23su 239Pu Day (MWd/tHM) (grams) (grams) (grams) I -I * .. I * -.. * ..

I -.. * .. * .. -.. * ..

Page 101of190 ATTACHMENT 1 Figure 66 Change ofl'rarget Assembly Power 5.3. 7 Summary of Target Assembly Nuclear Design Evaluation The nuclear performance of the TA has been evaluated TA model are as follows:

Two stand-alone target cartridges together are sufficiently subcritical, with a Keff below ... mrhe MURR driver fuel element power peaking factor is less than -mrhe variation of target pellet linear power and total TA power due to CB movement is small.

mrhe uncertainty due to pellet fabrication does not exceed

The uncertainty due to mechanical tolerance of the target rod loading -for the peak linear power of the extreme bumup core and total target power of the maximum bumup core, respectively. Page 102of190 ATTACHMENT 1 -Nuclear compatibility of the target cartridges and assembly with the MURR core has been assessed based on applicable MURR Technical Specifications. In general, because the thennal power from the

T As is relatively small -of the nominal MURR core thermal power) and the target rods are loaded in the graphite reflector region outside the Be reflector, the effect of target rods on the core reactivity characteristics is small as summarized below for representative core bumup states with the base target loading:

The least negative reactor core temperature coefficient ofreactivity is -1.69 x 10-4 .!lk/k/°C, which is more negative than Technical Specification 5.3.a value of -1.08 x 10-4 .!lk/k/°C (-6.0 x 10-5 .!lk/k/°F). The primary coolant temperature range in this calculation was from 20.44 °C (68.79 °F) to 76.84 °C (170.31 °F).

The reactor core void coefficient of reactivity is -3.4 x 10-3 .!lk/k/%void, which is more negative than Technical Specification 5.3.b value of -2.0 x 10-3 .!lk/k/%. The primary coolant density in the calculation was changed from 100% nominal to 0.1 %.

The highest regulating blade reactivity worth is 4.02 x 10-3 .!lk/k, which is less than Technical Specification 5.3.d limit of 6.0 x 10-3 .!lk/k.

The lowest subcriticality margin of the core with the most reactive shim blade and regulating blade fully withdrawn is 0.055 .!lk/k, which is greater than Technical Specification 3.1.e value of 0.02 .!lk/k.

The excess reactivity of the cold, clean minimum bumup core with cold, fresh, base TA loading is 0.072 .!lk/k, which is less than Technical Specification 3.1.a value of 0.098 .!lk/k referenced to the cold, clean critical core.

A single target cartridge insertion into the MURR graphite reflector regions adds -to the MURR core, which is lower than allowed reactivity worth of each secured removable experiment, 0.6% .!lk/k (Technical Specification: 3.1.a). 6. Target Assembly Thermal Hydraulic Design Analysis This section provides information that demonstrates that adequate cooling capacity is available to keep the TA in a thermally safe condition during all operational states. The TCS design is described in detail in Section 3. For 10 MWt reactor power, the peak linear power, including margin for uncertainties, is -which corresponds to a local target cartridge power density of-. The target rod with this power density is located in the middle of the cartridge, and has an average power density of-. -Page 103of190 ATTACHMENT 1 It has a peak linear power, including margins for uncertainty, of-' and an average power density of-' with a local maximum of Both the peak power and the maximum total heat cases were investigated in these analyses to ensure safe operations. In addition to the power generated in the target rods, (Table 17). This heat is also absorbed by the target cartridge rod cooling water and removed by the TCS. The maximum local target rod powers -are used to determine the local peak temperatures and heat fluxes; the cartridge power is used to determine the total coolant heat absorption and hence, flow rate and outlet temperatures. The pellet with the peak linear heat rate has that peak approximately

of the rod length above the coolant entrance, at a point where -of the target rod heat is generated. The target rod with the highest total power has its maximum power pellet approximately

of the rod length above the coolant entrance, at a point where -of the rod heat is generated. Analyses were performed for both target rods at their point of maximum heat generation rate, at the local water temperature and pressure. Margins were added to account for uncertainties in flow rate, pressure drop and temperature measurements. All analyses were performed assuming which is the burnup experienced for the maximum power density pellet (greater burnup rates increase the relocation of the pellet, with a resultant reduction in the pellet/clad gap, which results in colder. temperatures). It has no effect on the CHFR. Varying burnup rates were considered in the transient analyses performed on the structure. Steady-state, one-dimensional, axisymmetric analyses were performed at the location of the maximum U02 pellet power density. The effect of reactor power (10 MW1 nominal and 11.5 MW1 maximum), coolant flow rate ), cartridge channel diameter ( -) and the number of active target rods per cartridge investigated to verify that all system temperatures, as well as the CHFR, were in safe regimes. 6.1 Thermal-Hydraulic Design Basis 6.1.1 Normal Conditions of Operation The mechanical design, dimensions and tolerances for the target rods are given in Table 3. The nominal (cold) gap between the pellet and the cladding is -The minimum/maximum (cold) gap is -respectively. Page 104 of 190 ATTACHMENT 1 The target rods are placed in a -cartridge which allows "annular" flow axially along the length of the rod. The target rod spacing is -* center to center, and the nominal cartridge flow full" hole size is-* Thus, on each side of the rod, the "annulus" defining the flow along the target rod shall be open and intersecting the adjacent scallop over an angle of -for the nominal geometry. This gives a cartridge flow channel with a hydraulic diameter of-* For the minimum channel diameter of-, the hydraulic diameter will be-* For the maximum channel diameter of -' the hydraulic diameter will be-* (Channel diameter variations are the result of manufacturing processes as well as cartridge pressure during operation.) The target cooling water will flow through the cartridge at a flow rate that will have a nominal heat rise of -* For a cartridge with. target rods, a target water mass flow rate of -will produce a nominal fluid velocity of-as it flows through the section of maximum heat flux. In order to minimize the thermal impact on .the reactor pool, the nominal target coolant outlet temperature is set to be the same as the pool water temperature, nominally . Thus, the nominal target cooling water inlet temperature is . The secondary coolant system, which is common to the MURR primary and pool coolant systems and the SGE TCS, will ensure that even on the hottest summer days, the required target inlet temperature is achieved. 6.1.2 Subcooled Nucleate Boiling For both the peak linear power target rod and the maximum total power rod, the nominal heat flux at the surface of the -cladding will be approximately -' which will cause subcooled nucleate boiling at the cladding wall surface. To examine the boiling effects and ensure safe target rod operation, multiple boiling convection models using the ANSYS FLUENT computational fluid dynamics (CPD) computer code were developed (Reference 23). In addition, a literature search was carried out to find relevant experiments to validate the analyses. The closest case found in the literature is that of Del Valle and Kenning (Reference 24) who studied subcooled nucleate boiling on flat plates under forced convection. For their experiments, they utilized water at 116.7 kPa (16.9 psia), with subcooling between 24 and 84 °C, and coolant velocities between 0.8 and 2.0 m/sec, at heat fluxes up to 4.6 MW/m2. , while the coolant pressure and velocity are lower. Therefore, their results should be conservative. Del Valle and Kenning's data show heat flux rising faster than other boiling correlations, such as Lottes and Chen (Reference 25). The Jens-Lattes correlation, used in the following thermal analysis, is an empirical correlation for fully-developed subcooled nucleate boiling, whereas the Chen model is a broader correlation which combines single phase forced convection and surface boiling effects. The Chen model is used in RELAP5 for nuclear thermal-hydraulic calculations. Del Valle and Kenning give several important observations from their photographic study of the bubble formation in their experiment. The photographic study was done for the 84 °C subcooling, 1. 7 mis coolant velocity condition. These are summarized below: Page 105of190 ATTACHMENT 1

At high subcooling, the flow remains in a bubbly regime up until burnout. "All runs at 84 °C subcooling remained in the bubbly flow regime up to burnout, the bubbles retaining their identities except for occasional coalescences leading to small, short-lived vapour patches."

The bubbles formed are small. "The maximum diameters were normally distributed (class interval 0.1 mm) with a mean value of0.4 mm (+/-5%) independent of heat flux."

The bubbles are short-lived. "Even at the maximum camera speed [of 10,000 frames per second], individual bubbles appeared in only 3 -5 frames." These observations indicate that while bubble nucleation is expected, the bubbles are small and collapse quickly. This agrees with an earlier publication by Unal (Reference 26), who writes, "The bubbles formed at high subcooling do not leave the heated surface! although they attain their maximum diameter. They slide along the heated surface or collapse, as observed visually by Gunther." Therefore, it is expected that while some subcooled nucleate boiling will occur in the high heat flux regions of the target section, there will be minimal vapor generation, and given the expected bubble lifetime and high subcooling, there will be no vapor present in the coolant exiting the target. The correlations above do not make predictions on vapor formation. To corroborate the work of Del Valle and Kenning, ANSYS FLUENT CFD models were used to examine the presence of vapor in the target coolant. These models used an Eulerian multiphase model to track the balance of liquid water and vapor throughout the fluid domain of the target housing. The heat transfer at the wall was determined by what is known as the "RPI model," developed at Rensselaer Polytechnic Institute (RPI). The RPI model partitions the heat flux into boiling and convective effects, much like the Chen correlation (Reference 27). The results from these CFD models were in line with the observations of Del Valle and Kenning, in that the presence of vapor was confined to the cladding rod surfaces and that the total amount of vapor generated was very small. The resulting vapor volume fractions at the maximum heat flux rod in the worst case operating conditions are shown in Figure 67. The graph shows that the maximum local vapor volume faction is about 5 parts per million. This is lower than the 100 parts per million shown in Del Valle and Kenning's experiments, although the difference is reasonable given that the coolant pressure and velocity are higher in the target (the latter by nearly a factor of 3). Additionally, the results show that all vapor bubbles collapse outside of the high heat flux region at the rod surface, and water exiting the target housing contains no vapor. Page 106of190 5.00E-06 I c 0 4.00E-06 .s 3.00E-06 GI E ::s g 2.00E-06 1.00E-06 O.OOE+OO 0.1 0.2 ATTACHMENT 1 -.-Front path -.-Back path -.-Side path 0.3 0.4 0.5 0.6 0.7 0.8 Axial position (m) Figure 67 Vapor Fraction at Cladding Wall for Worst-Case Conditions in FLUENT RPI Wall Boiling Model Both the CFO modelling and Del Valle and Kenning's work give compelling evidence to show that vapor formation in the target is minimal and confined only to the highest heat flux regions of the target rods. To confinn this evidence, experiments are planned to replicate the exact conditions that the target rods will experience in the MURR graphite reflector region, and provide data regarding CHF and subcooled nucleate boiling behavior, which should confirm that two-phase flow is not a concern in the target. The TA thermal design and analysis ensures that vapor bubbles do not coalesce together forming a vapor film at the cladding surface. This phenomenon occurs when the surface heat flux reaches the CHF value at the transition from nucleate boiling to partial film boiling (Reference 28). Three correlations, Bernath (Reference 29), Macbeth (Reference 30) and Groeneveld (Reference 31) have been used to assure that the target operates in a safe heat transfer regime. The Bernath correlation is very conservative (in predicting low critical heat fluxes) and is an industry-wide standard for such The Macbeth correlation incorporates data at the low pressures of this system, and it is one of the correlations recognized in the NRC light water reactor fuel rod analysis code FRAPTRAN. The Groeneveld correlation, a lookup table based on the most extensive set of data available, is generally considered the most accurate. All three (3) predictions were found to be in close agreement. The lowest CHF of the three (3) correlations (in every case, Bernath) was used to calculate the CHFR, defined as the CHF divided by the actual heat flux. Recognizing the large amount of scatter and uncertainty in CHF data, a CHFR > 2.0 is required and has been assigned for the design. With a subcooling margin greater than 80 °C, the TA is normally operating at the transition point from convection to nucleate boiling, well below the value of heat flux where transition from nucleate to partial film boiling occurs. Page 107of190 ATTACHMENT I The thermal analysis first calculates the single phase convective heat transfer from the cladding surface. If the surface temperature is higher than that required to initiate subcooled nucleate boiling per the Jens Lottes correlation (Reference 25), then the water at the surface is assumed to be boiling. Its surface temperature is set at the Jens-Lottes suggested value, and the CHFR is calculated. Figure 68 shows the CHFR, heat flux and enthalpy as a function of the axial location in the target rod for both the peak linear power (PLP) rod and the maximum heating (MH) rod. a: LL ::c u Axlal Location, cm Figure 68 >-.e-ca .s= -c w -c ca 0 0 u ... 0 >< :J u::: -ca Cl> ::c LEGEND -*-*-*-PLP He* Flu .. MWm't -*-*-*-MH He* FIW<, MWlm't PLP Bernath CHFR MH Berna!h CHFR PLP Groeneveld CHFR MH Groeneveld CHFR PLP Macbeth CHFR MH Macbeth CHFR PU' Qod.,i Wwlpy MJ'9t MHCcGlanl Enhl"1, ""'vt t -uses ri!tit Y axis Heat Flux and CHFR as a Function of Axial Location for Peak Power Target Rods The red dotted line shows the CHFR (Bernath) for the peak linear power target rod. The blue dotted line shows the CHFR (Bernath) for the maximum power rod. In all these analyses, the Bernath CHFR was consistently the most conservative. Although maximum heat flux occurs at different points along the PLP and MH rods, the minimum CHFR is almost the same for both rods. The small temperature rise of the coolant yields an almost constant line for the coolant enthalpy as a function of rod axial location. Thus, the point of the minimum CHFR coincides with the location of the maximum heat flux in the rod. 6.1.3 Design Margins for Uncertainties In addition to a safety margin of. on heat transfer coefficients, several additional margins to account for operating measurement uncertainties were included in these analyses: Page 108of190 ATTACHMENT 1 1. The pressure drops through the entire TA that were calculated using CFD were reduced by

for the thennal analyses as this will reduce the water saturation temperature and give a lower predicted CHFR. 2. A control system uncertainty of-was applied to the cooling water inlet temperature. As with the pressure drop reduction, this reduces the predicted CHFR. 3. An uncertainty of. was applied to the system flow rate to account for possible errors in measurement -this also reduces the predicted CHFR. Multiple analyses were performed to investigate the possible variations of operating parameters and geometric tolerances. Each analysis was performed for both the nominal parameters and with the above margins for operating uncertainties. The first set of results represents the most accurate possible engineering solution of the case analyzed. The second set ofresults represents the worst possible stacking of all the operating uncertainties. The thermal design is required to have a CHFR > 2.0 for the combined application of uncertainties for the worst possible operating condition. A CHFR of 2.0 is a 100% margin by itself. 6.1.4 Target Assembly Steady-State Operations The water flow rate to the TA has a value of below the pool temperature. It is conservatively assumed that

of the flow leaks out at the labyrinthine seal at the junction of the removable cartridge and the target housing. This leakage will be determined by ex-reactor testing on the unit. The target flow rate is therefore . After flowing through the cartridge, it exits to the reactor pool at a pool depth of . The absolute pressure at the diffuser exit is -and the pressure of the water at the mid-height of the target (location of maximum heat flux) is approximately . This sets the local water saturation temperature at The nominal water inlet and outlet temperatures are , respectively. This corresponds to a reactor pool temperature of temperature rise corresponds to target rods in the cartridge, each dissipating an average of -* During ramp up to full scale operation, between rods will be present in the cartridge, with the balance occupied by non-heat generating filler rods. In order to maintain the TA water outlet temperature equal to the pool temperature, the TA inlet temperature is automatically adjusted so that: Tout,mixed = TrooL = [nx(Tinlet +lO)+(ll-n)xTin1e1]/ll. Hence, Tinlet = Tout, mixed -10/11 x n = TrooL -0.9091 x n. Table 28 shows the variation in coolant temperatures as the active rod count varies. Page 109of190 ATTACHMENT 1 Table 28 Variation of Coolant Inlet Temperature with Active Target Rod Count Active Non Heating Tin T;u Tout Tout Mixed Tout Mixed Tout Rods Rods (QC) (°F) (QC) (oF) (QC) (°F) I I .. .. .. -.. -I I .. .. .. -.. -I I .. Im .. -.. -I I .. .. .. -.. -I I .. .. .. -.. -I I .. .. .. -.. -I I .. .. .. -.. -* I .. .. .. -.. -* I .. .. .. -.. -To achieve a cartridge mixed outlet temperature equal to a nominal pool temperature of the range for T inlet will be For each additional degree of initial temperature rise of the pool, the cartridge inlet temperature will need to increase by the same number of degrees. For these analyses, it was assumed that -active target rod operation would not occur if the pool temperature exceeded This limit is dictated by the requirement to maintain CHFR's above 2.0 for all operational scenarios. As can be seen from Table 28, both the inlet and outlet coolant temperatures rise with decreasing active target rod counts, the most severe operating conditions are those with the least number of active rods. As the system heats up, the gap will be set by the relative expansion of the pellet and the cladding, as well as the physical expansion -because of relocations in the pellet due to the stresses experienced during fuel bumup. Changes in the gap also change the volume occupied by the helium between the pellet and the cladding, which, along with temperature changes, increase the pressure of the gap helium. It is noted that the thermal conductivity I -is a function of temperature, but in very small gaps such as these, it is also a function of the gap size and pressure. Properties used for these analyses are shown in Figure 69, Figure 70, and Figure 71. Page 110of190 20 18 0 16 E ;: 14 -> 12 ;: (,) ::::J ,, 10 c 0 8 0 ftS E 6 ... 1! .... 4 2 ATTACHMENT 1 .E §_ c 0 ii c a. >< ...................................................... w ftS E ... 1! .... 0 c: u =i 8 500 1000 1500 2000 Temperature, °C Figure 69

Thermal Conductivity and Thermal Expansion Coefficient 20 20 0 e .._ 18 18 E ::l 0 16 16 c E 0 ii ;: c 14 14 ftl i-Q. 2! 12 12 -ftl (,) E ::::J 10 10 ,, .! c 0 8 8 .... 0 -ftl 0 ....... --........ u ....... ................. , ......... -E 6 .............................. 6 c ... ............ .!! .! (,) .... 4 4 = 8 2 2 0 .!!! 00 "Cl 100 200 300 400 ftl Temperature, °C a: Figure 70 LEGEND Thermal Cond.Jdnnty CTEt t *uses Y axis LEGEND Thermal O:uicb::tlvlty Radial CTEt t *uses ri!tit Y axis -Thermal Conductivity and Radial Thermal Expansion Coefficient Page 111 of 190 CJ 0.40 E i > 0.:ll -u :I 'a c 0 0.20 CJ ca E ... ! ._ 0.10 ATTACHMENT I ............ .. ****** , ................... .. ............. .......... ............. ............. 100 200 300 400 500 600 700 800 900 1000 Temperature, °C Figure 71 LEGEND -Thermal Conductivity in Small Gaps Target rod thennal analyses were performed for dimensional tolerances 5 µmgap 10µm gap 20µm gap 40µm gap , nominal and maximum reactor power (I 0 and 11.5 MW) and coolant flows (nominal flow rate value of 100% and the SCRAM limited flow rate value of active target rods and for both the middle peak power rod and the end rod with the maximum linear heating. The case for -rods is important because it has the highest inlet temperature thus reducing the margin to saturation. This may be seen in the slightly lower CHFR's. Of all the cases investigated, the smallest value of the CHFR was seen in the case of the peak power target rod; with nominal pellet/clad geometries and the largest possible water channel. In all cases investigated, the CHFR was greater than or equal to 2.0. Table 29 shows the details of steady-state operations for , for both the nominal and worst case uncertainties, for both the peak power and maximum heating rods at 10 MW reactor power with 100% TCS flow. Table 30shows the details of those same cases but at 11.5 MW, reactor power and

TCS flow. This is the worst possible case (as the reactor will SCRAM at either higher power or lower flow). Table 31 and Table 32 show the corresponding three (3) target rod cases. Table 33 shows the worst (lowest CHFR) cases, those for -target rods, with a -gap and added calculational margin. The CHFR for this condition is 2.07. By comparison, the expected minimum CHFR is 2.92 per Table 29. Page 112of190 ATTACHMENT 1 Table 29 Predicted Thermal Performance for -Target Rods, 10 MWt Reactor Power, -Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Maximum Heating Rod Heating Rod, Margin Added Margin No Added Maximum Mar2in Added Mar2in Reactor power, MW1 * * *

Flow% * * *

Active rods * * *

Pool temperature, °C * * *

Target inlet temperature, °C * * *

Cold gap ----Rod type ----Rod location --*

Added error margin .. -.. -I-----Cold gap, µm * * * *

radial growth, µm .. .. * .. Cladding radial growth, µm * * *

Relocation growth, µm * * *

Hot gap, µm * * * * -gap pressure, atm * * *

Plenum height, mm (Cold) * * *

Plenum height, mm (Hot) .. .. .. .. Axial CTE-µm/(m°C) * * *

Radial CTE -µm/(m°C) .. .. .. .. Radial CTE -' µm/(m°C) * * * *

centerline temperature, °C ----* average temperature, °C ----* surface temperature, °C .. .. .. .. -temperature, °C .. .. .. .. Cladding temperature at ID, °C .. .. .. .. Mean cladding temperature, °C .. .. .. .. Cladding temperature at OD, °C .. .. .. .. Page 113of190 ATTACHMENT 1 Table 29 Predicted Thermal Performance for -Target Rods, 10 MWt Reactor Power, -(continued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Maximum Heating Rod Heating Rod, No Added Maximum Margin Added Margin Margin Added Margin

gap conductance, W l(m2 C) ----Coolant HTC, Wl(m2 C) ----Water-OD, mm .. -.. -Dh,mm --* .. Re (D11) ----Local coolant velocity, mis * *I *

Mass flow per rod, kgls ----Mass flow per target, kg/s * -* -Mass flow per target, gpm ----Coolant inlet temperature, °C .. .. .. .. Local coolant temperature, °C .. .. .. .. Coolant outlet temperature, °C .. .. .. .. Local saturation temperature, °C ----Peak pellet power density, Wice ----Heat (modeled) per rod, W ----Heat flux at cladding OD, Wlm2 ----Bernath CHF, Wlm2 ----BemathCHFR * * *

Macbeth CHF, Wlm2 ----Macbeth CHFR * * *

Groeneveld CHF, W lm2 ----Groeneveld CHFR * * *

Minimum CHFR .. .. .. .. Pressure at CHFR, atm ----Page 114of190 ATTACHMENT 1 Table 30 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, -Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Maximum Heating Rod No Heating Rod, Margin Added Margin ;\.dded Margfo M.aximum Added Mar2in Reactor power, MW1 * * *

Flow% * * *

Active rods * * *

Pool temperature, °C * * *

Target inlet temperature, °C * * *

Cold gap ----Rod type ----Rod location --*

Added error margin .. -.. -Scallop diameter ----Cold gap, µm * * * * -radial growth, µm .. .. .. .. Cladding radial growth, µm * * *

Relocation growth, µm * * *

Hot gap, µm * * * * -gap pressure, atm * * *

Plenum height, mm (Cold) * * *

Plenum height, mm (Hot) .. .. .. .. Axial CTE *. µm/(m0C) * * *

Radial CTE *. µm/(m°C) .. .. .. .. Radial CTE -' µm/(m0C) * * * *

centerline temperature, °C ----* average temperature, °C ----* surface temperature, °C .. .. .. .. -temperature, °C .. .. .. .. Cladding temperature at ID, °C .. .. .. .. Mean cladding temperature, °C .. .. .. .. Cladding temperature at OD, °C .. .. .. .. Page 115of190 ATTACHMENT 1 Table 30 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, (continued) Peak Power Peak Power Max Linear Max Linear Rod, No Added . Rod, Maximum Heating Rod No Heating.Rod, Maximum Margin Added Margin Added Margin Added Margin

gap conductance, W /(m2 C) ----Coolant HTC, W/(m2 C) ----Water-OD, mm ----D11,mm --.. .. Re (D11) ----Local coolant velocity, mis * * *

Mass flow per rod, kg/s ----Mass flow per target, kg/s ----Mass flow per target, gpm .. .. .. .. Coolant inlet temperature, °C .. .. .. .. Local coolant temperature, °C .. .. .. .. Coolant outlet temperature, °C .. .. .. .. Local saturation temperature, °C ----Peak pellet power density, W /cc ----Heat (modeled) per rod, W ----Heat flux at cladding OD, W/m2 ----Bernath CHF, W/m2 ----BemathCHFR * * *

Macbeth CHF, W /m2 ----Macbeth CHFR * * *

Groeneveld CHF, W /m2 ----Groeneveld CHFR * * *

Minimum CHFR .. .. .. .. Pressure at CHFR, atm ----Page 116of190 ATTACHMENT 1 Table 31 Predicted Thermal Performance for -Target Rods, 10 MWt Reactor Power, -Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin Added Margin Margin Added Margin Reactor power, MW1 * * *

Flow% * * *

Active rods I I I I Pool temperature, °C * * *

Target inlet temperature, °C * * *

Cold gap ----Rod type ----Rod location --*

Added error margin .. -.. -Scallop diameter ----Cold gap, µm * * * * -radial growth, µm .. .. * .. Cladding radial growth, µm * * *

Relocation growth, µm * * *

Hot gap, µm * * * * -gap pressure, atm * * *

Plenum height, mm (Cold) * * *

Plenum height, mm (Hot) .. .. .. .. Axial CTE *. µm/(m°C) * * *

Radial CTE *. µm/(m0C) .. .. .. .. Radial CTE -' µm/(m0C) * * * *

centerline temperature, °C ----* average temperature, °C ----* surface temperature, °C .. .. * .. -temperature, °C .. .. .. .. Cladding temperature at ID, °C * .. .. .. Mean cladding temperature, °C * .. .. .. Cladding temperature at OD, °C .. .. .. .. Page 117 of 190 ATTACHMENT 1 Table 31 Predicted Thermal Performance for -Target Rods, 10 MWt Reactor Power, -(continued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin Added Mar2in Mar2in Added Mar2in

gap conductance, W/(m2 C) ----Coolant HTC, W/(m2 C) ----Water-OD, mm ----D1i,mm --.. .. Re (D1i) ----Local coolant velocity, mis * * *

Mass flow per rod, kg/s ----Mass flow per target, kg/s ----Mass flow per target, gpm ----Coolant inlet temperature, °C .. .. .. .. Local coolant temperature, °C .. .. .. .. Coolant outlet temperature, °C .. .. .. .. Local saturation temperature, °C ----Peak pellet power density, W /cc ----Heat (modeled) per rod, W ----Heat flux at cladding OD, W/m2 ----Bernath CHF, W/m2 ----Bernath CHFR * * *

Macbeth CHF, W/m2 ----Macbeth CHFR * * *

Groeneveld CHF, W/m2 ----Groeneveld CHFR * * *

Minimum CHFR .. .. .. .. Pressure at CHFR, atm ----Page 118of190 ATTACHMENT 1 Table 32 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, -Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum . No Added Maximum Margin Added Margin Margin Added Margin Reactor Power, MW1 * * *

Flow% * * *

Active rods I I I I Pool Temperature, °C * * *

Target inlet temperature, °C * * *

Cold Gap ----Rod Type ----Rod location -* *

Added Error Margin .. -.. -Scallop Diameter ----Cold Gap, µm * * * * -radial growth, µm .. .. .. .. Cladding radial growth, µm * * *

Relocation Growth, µm * * *

Hot Gap, µm * * * * -Gap pressure, atm * * *

Plenum height, mm (Cold) * * *

Plenum height, mm (Hot) .. .. .. .. Axial CTE ., µm/(m0C) * * *

Radial CTE ., µm/(m°C) .. .. .. .. Radial CTE -µm/(m°C) * * * *

centerline temperature, °C ----* average temperature, °C ----* surface temperature, °C .. .. .. .. -temperature, °C .. .. .. .. Cladding temperature at ID, °C .. .. .. .. Mean cladding temperature, °C .. .. .. .. Cladding temperature at OD, °C .. .. .. .. Page 119of190

.--------------------------------------------------ATTACHMENT 1 Table 32 Predicted Thermal Performance for -Target Rods, 11.5 MWt Reactor Power, -(continued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin .Added Margin Man?:in Added Margin

gap conductance, W l(m2 C) ----Coolant HTC, Wl(m2 C) ----Water (scallop) OD, mm ----D1i,mm -.. .. .. Re (D1t) ----Local coolant velocity, mis * * *

Mass flow per rod, kgls ---* -Mass flow per target, kg/s ----Mass flow per target, gpm .. .. .. .. Coolant inlet temperature, °C .. .. .. .. Local coolant temperature, °C .. .. .. .. Coolant outlet temperature, °C .. .. .. .. Local saturation temperature, °C ----Peak pellet power density, Wice ----Heat (modeled) per rod, W ----Heat flux at cladding OD, Wlm2 ----Bernath CHF, Wlm2 ----Bernath CHFR * * *

Macbeth CHF, W lm2 ----Macbeth CHFR * * *

Groeneveld CHF, Wlm2 ----Groeneveld CHFR * * *

Minimum CHFR .. .. .. .. Pressure at CHFR, atm ----Page 120of190 ATTACHMENT 1 Table 33 Predicted Thermal Performance for -Target Rods, Worst-Case Operations Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin Added Margin Margin Added Margin Reactor power, MW1 * * *

Flow% * * *

Active rods I I I I Pool temperature, °C * * *

Target inlet temperature, °C * * *

Cold gap ----Rod type ----Rod location -* -* Added error margin ----Scallop diameter ----Cold gap, µm .. .. .. .. -radial growth, µm .. .. * .. Cladding radial growth, µm * * *

Relocation Growth, µm * * *

Hot gap, µm * * * * -gap pressure, atm * * *

Plenum height, mm (Cold) * * *

Plenum height, mm (Hot) .. .. .. .. Axial CTE *. µm/(m0C) * * *

Radial CTE *. µm/(m0C) .. .. .. .. Radial CTE -* µm/(m°C) * * * *

centerline temperature, °C ----* average temperature, °C .. .. --* surface temperature, °C .. .. .. .. Helium temperature, °C .. .. .. .. Cladding temperature at ID, °C .. .. .. .. Mean cladding temperature, °C .. .. .. .. Cladding temperature at OD, °C .. .. .. .. Page 121 of 190 ATTACHMENT 1 Table 33 Predicted Thermal Performance for -Target Rods, Worst-Case Operations (continued) Peak Power Peak Power Max Linear Max Linear Rod, No Added Rod, Heating Rod Heating Rod, Maximum No Added Maximum Margin Added Margin Margin Added Margin

gap conductance, W/(m2 C) ----Coolant HTC, W/(m2 C) ----Water (scallop) OD, mm ----D1i,mm -.. -.. Re (D1i) ----Local coolant velocity, mis * * *

Mass flow per rod, kg/s ----Mass flow per target, kg/s ----Mass flow per target, gpm --.. .. Coolant inlet temperature, °C .. .. .. .. Local coolant temperature, °C .. .. .. .. Coolant outlet temperature, °C .. .. .. .. Local saturation temperature, °C ----Peak pellet power density, W /cc ----Heat (modeled) per rod, W ----Heat flux at cladding OD, W/m2 ----Bernath CHF, W/m2 ----Bernath CHFR * * *

Macbeth CHF, W/m2 ----Macbeth CHFR * * *

Groeneveld CHF, W/m2 ----Groeneveld CHFR * * *

Minimum CHFR .. .. .. .. Pressure at CHFR, atm ----Page 122 of 190 ATTACHMENT 1 6.1.5 Flow Induced Vibrations An analysis of the displacement of target rods caused by vibration due to the surrounding flow was perfonned. Ignoring the stiffening effects of the tube filled with the target pellets, the fixed-end -I tubes have a natural frequency of .. and a maximum displacement of-, using the correlations of Paidoussis, which is also the ASME B&PV recommended method (Reference 32). A maximum displacement of-is calculated using Farmer et.al (Reference 33). In either case, it can be concluded that vibrations are not an issue for the target systems. 6.1.6 Cooling of the Beryllium Reflector during Operations with Target Rods The Be reflector rests between the TA and the outer reactor pressure vessel. Nuclear, fluid flow and heat transfer analyses were perfonned to show the effect of the T As in the two (2) graphite reflector positions. The important consideration is the increased heating of the Be reflector and the effect on cooling of the Be surfaces. The MURR Be reflector generates up to 200 kW of heat, primarily due to neutron scattering and absorption of high energy photons from the reactor core. With the target rods installed, the MCNP analyses show that the

targets rods will increase the reflector power density by as much as

at the reactor core elevation centerline (its hottest point). This heat is then rejected to the MURR pool coolant system. It is not possible to measure the flow velocity around the Be reflector. Therefore, a model of the reflector cooling flow was created that accounts for these distinct flow paths. 1. Through the flux trap inside the annular core pressure vessel, which has a minimal effect on the reflector cooling. 2. Flow that travels between a 0.56-inch (14.2 mm) gap between the outer reactor pressure vessel and the inner wall of the Be reflector. The reactor regulating and control blades also reside in this gap and were accounted for. 3. The flow that passes through the gaps around the graphite reflector elements (and TAs) and down through holes and slots in the reflector support plate. A portion of this flow contacts the outer wall of the beryllium reflector, and thus is relevant to the reflector cooling. Measurements show the pressure drop across the reflector assembly is , which will serve as the basis for determining the flow rate in the reflector region. The flows through the three (3) flow paths were iterated in the model until all flow paths had the measured pressure drop and the total flow equaled the measured flow. The flow surrounding the Be reflector, the predicted power density in the reflector, and the resulting heat transfer were examined to determine if the pool coolant system provides adequate cooling to the Be reflector in the presence of two T As installed in graphite reflector positions (Reference 34). Worst-case control/regulating blade positions, reflector age and estimated power densities were used. The results show that the reflector maintains a 20 °C (36 °F) margin from surface temperature to coolant saturation temperature in the presence of. T As and hence there is no impact on Page 123 of 190 ATTACHMENT 1 cooling of the Be reflector. Additionally, the driving pressure head was shown to have a small effect on the flow rate, indicating that it is unlikely the surface temperature of the Be will increase significantly if the pressure head changed. 7. In-Pool Target Transfer System 7.1 Cartridge Loading/Unloading Station Design This section presents a description of the equipment and process that will take place to move target rods from/to target storage to/from the target cartridge at the cartridge loading/unloading station. 7.1.1 Cartridge Loading/Unloading Station Description The cartridge loading/unloading station is located in the weir area of the reactor pool. It is an aluminum structure designed to bolt to the existing pool structure and contains storage for

target rods, -target cartridges, and -target diffusers. Figure 72 shows the location of the in-pool storage area within the MURR reactor pool and how it is positioned in relation to the reactor. CARTRIDGE LOADl1'"G/UlloLOADl1'"G Figure 72 Target Rod Loading/Unloading/Storage Location The cartridge loading/unloading station has three functions (Figure 73): I. It provides a location for the removal and installation of the diffuser from the cartridge; Page 124 of 190 ATTACHMENT 1 2. It holds the cartridge while target rods are removed or installed into the cartridge; and 3. It provides an approved target rod storage location that places the target rods in a geometry to ensure a large margin to criticality. Figure 73 Representation of the Cartridge Loading/Unloading Station Page 125of190 ATTACHMENT 1 During nonnal operations, no more than

target rods will be either in a cartridge or in target storage at the cartridge loading/unloading station. Of those. target rods, only. of them will be fresh. However, the performed criticality analysis assumes

target storage positions and -nearby target cartridges with

positions each filled with fresh target rods for a total of. fresh target rods in the station at the same time. The criticality analysis shows a maximum Keff of 0.57 under all conditions of moderation and reflection. MURR Technical Specification 5.4.a requires that fueled devices outside the reactor core be stored such that the Keff is less than 0.9 under all conditions of moderation and reflection. Administrative controls similar to the MURR administrative controls for reactor fuel handling will be implemented to ensure that the location of any target rod is known at all times and that the target rods are placed in the correct irradiation and storage locations. 7.1.2 Cartridge Loading/Unloading Station Operation The normal sequence of operation at the cartridge loading/unloading station will be: 1. Lower -fresh target rods into the station storage location. 2. Insert -fresh target rods into a cartridge at the station using the remote handling tool shown in Figure 71. 3. Bring an irradiated target cartridge from the target housing. 4. Remove the diffuser from the cartridge and place into storage basket. 5. Remove irradiated target rods from the irradiated cartridge and place them into the station storage location. 6. Move target rods as necessary to complete the placement of the next

target rods to be irradiated. (Note: The next. target rods to be irradiated may be all fresh or may be a mixture of fresh and previously irradiated target rods.) 7. Move the diffuser from the storage basket and install it onto the target cartridge to be irradiated. 8. Move the loaded target cartridge to be irradiated into the target housing at the reactor reflector (target rod remote handling tool shown in 9. Figure 74). Page 126of190 ATTACHMENT 1 Figure 74 Target Rod Remote Handling Tool 7.2 Installation and Removal of the Target Cartridge Into/From the Target Housing 7.2.1 Target Cartridge Description A detailed description for the target cartridge and how it fits into the target housing is provided in Section 2.1.4. 7.2.2 Cartridge Installation and Removal Into/From the Target Housing Operation The normal sequence of operation for transferring the irradiated target cartridge from the target housing to the cartridge loading/unloading station will be: 1. Shut down the reactor. 2. To allow sufficient cooling and decay, keep the target cartridge in the target housing for a minimum of one (1) hour after reactor shutdown prior to unlatching the cartridge. 3. Remove the RTD at the top of the diffuser and secure it out of the way of cartridge movement operations. 4. Thread the cartridge removal tool into the top of the diffuser. 5. Unpin the cartridge by turning bolt-connected levers with a socketed, long tool. Page 127 of 190 ATTACHMENT 1 6. Lift the cartridge vertically with the attached cartridge removal tool. 7. Move the cartridge to the cartridge loading/unloading station. 8. Once the cartridge is attached to the station, unthread the cartridge removal tool. The nonnal sequence of operation for transferring the target cartridge from the cartridge loading/unloading station to the target housing will be: I. 2. 3. 4. 5. 6. 7. 8. 9. 8. Thread the cartridge removal tool into the top of the diffuser Move the cartridge to the just above the target housing in which the cartridge will be irradiated. Slowly lower the cartridge into the housing making sure the cartridge rail features engage the target housing rails as seen in Figure 5. Gently push downward to ensure cartridge is resting in the target housing with proper alignment pin engagement. Secure the cartridge by turning bolt-connected levers with a socketed, long tool until no further movement is detected. Pull upward and push downward on the attached cartridge removal tool to ensure the cartridge is secured and cannot move. Visually observe the proper position of the cartridge locking mechanism. Unthread the cartridge removal tool. Install the RTD at the top of the diffuser. Radiological Protection Evaluation for the SGE Target Experimental Facility Operations This section discusses and analyzes the expected radiological consequences related to normal operations from the production of Mo-99 using the SGE TEF. Included are the principal discussions of the MURR facility program to control radiation and expected radiation exposures due to operation, maintenance, and use of the irradiation hardware associated with the TEF. This section also outlines the methods for quantitative assessment of radiation doses in the restricted and unrestricted areas; application of these methods to all applicable radiation sources related to the operation of the TEF; and the program and provisions for protecting the health and safety of all individuals present at MURR, the general public and the environment. This section does not discuss the Radiation Protection Aspects of the MURR Health Physics program as related to the processing of the LEU targets for the Selective Gaseous Extraction (SGE) of Mo-99 from the TEF. That will be covered under the Part 2 License Amendment submission for processing the LEU target rods and the extraction ofMo-99. The MURR Radiation Protection Program (RPP) has been established to protect the health and safety of all individuals present at MURR, the general public and the enviromnent. In accordance with 10 CFR 20.1101, this program has been developed, documented, and implemented to a level commensurate with the scope and extent of licensed activities at MURR, and is sufficient to ensure compliance with the Page 128of190 ATTACHMENT 1 regulations in 10 CFR 20. A primary component of this program is the fundamental principle of maintaining individual radiation exposures and releases of radioactive effluents as low as is reasonably achievable (ALARA). Responsibility for maintaining the MURR ALARA Program extends to all individuals who are granted access to the reactor facility. Renewed Facility Operating License No. R-103 is the primary license that covers the authority and responsibilities associated with the reactor and the majority of radioactive materials produced at MURR. It is under the auspices of this license that operation of the TEF will occur. Radioactive Material use is further supplemented by the authorization of an NRC-issued Broad Scope license (24-00513-39) and is used to support the research and development mission of the MURR for materials and activities which may not be currently produced or covered under license No. R-103. The radiation sources that are expected to be generated during irradiation of the TEF fall under the administrative control of the MURR Radiation Protection and Radioactive Waste Management Programs and can be categorized as airborne, liquid, or solid. While each of these categories is discussed individually in the following paragraphs, the major contributors to each category can be summarized as follows: 8.1 Airborne Sources Airborne -Potential airborne sources consists mainly of the fission product gasses of the bromine, iodine, krypton and xenon species. Of these, iodine-131 (I-131) has the lowest Derived Air Concentration (DAC) and Effluent Release limits due to potential thyroid exposure while krypton-85 (Kr-85) has the longest half-life (10.76 years); however, Kr-85 is a noble gas and as such is minimally active with regards to human and non-human biological effects. All potential airborne sources from the SGE TEF are confined by the target rod cladding and are not expected to be released during routine operations. Currently, during normal operation of MURR, argon-41 (Ar-41) is the principal source of airborne radioactivity (> 99%) released through the facility ventilation exhaust stack. However, similar to the irradiation of reactor fuel, there exists the possibility to release any one of the fission products created during the fission process from the TEF. Such release would likely be limited to noble gases (krypton and xenon) and the halogens iodine and bromine that exist in the "gap" between the LEU pellets and the target cladding. These releases could occur during two (2) points of the irradiation and use of the targets; (1) during the actual irradiation of the target rods in the graphite reflector region (this would occur if a leak occurs in the cladding of the target rod, thereby allowing the gap fission gases to escape), and (2) during handling of the target rods in the reactor pool pre-and post-irradiation. 8.2 Liquid Sources Liquid -Liquid sources would include primary and pool coolant used in the reactor coolant systems and the TEF target cooling water, which is essentially reactor pool water that has been pumped to a separate cooling system to facilitate target heat removal. The water becomes activated and contaminated by direct neutron activation of the water molecule itself (H-3) along with any activation of impurities contained in the water due to its use as a coolant. The pool water would also contain contaminants due to its direct Page 129of190 ATTACHMENT 1 contact with the reactor and pool coolant system components. Since primary and pool coolant is, by design, contained to the maximum extent possible, there are no substantial releases of these liquids directly into the environment. However, certain reactor maintenance activities along with TEP cooling system maintenance could result in small volumes of liquid (containing mainly tritium) being directed to the existing liquid waste retention system. Limited and strictly controlled quantities of liquid radioactive waste are released to the sanitary sewer in accordance with the requirements of 10 CPR 20.2003; this effluent may contain small quantities of pool and primary water. Thus the operation of the TEP would add little to existing liquid waste effluent streams. All potentially radioactive aqueous liquid wastes are directed to a liquid waste retention and disposal system located at a level below the main grade of the MURR laboratory building. Liquid waste is then retained or mechanically processed until an assay indicates that activity concentration levels are less than the limits specified in 10 CPR 20, Appendix B for disposal by release into sanitary sewerage. Tritium currently accounts for about 81 % of the total activity released each year. Historically, H-3 release is < 5% of the annual release limit of 5 Ci. Pool coolant is the only radioactive liquid source expected to be impacted by the TEF. Radioactivity in this liquid occurs by the same mechanisms as described above: neutron interactions with hydrogen and oxygen in the water and neutron interactions with system and structural components, with the subsequent transfer into the pool coolant. Table 34 contains a list of the predominant radionuclides and their typical measured concentrations present in the pool coolant at 10 MW. Table34 Predominant Radionuclides in the MURR Pool Coolant and Their Measured Concentrations at 10 MW Radionuclide Half-Life Typical Concentration1 Magnesium-27 (27Mg) 9.45 minutes 1.1 O x 10'2 Ci/ml Sodium-24 (24Na) 14.96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 4.64 x 10-3 Ci/ml Manganese-56 (56Mn) 2.58 hours6.712963e-4 days 0.0161 hours 9.589947e-5 weeks 2.2069e-5 months 2.54 x 10-3 Ci/ml Technetium-101 (101Tc) 14.2 minutes 4.70 x 10-5 Ci/ml Technetium-99m (99nvr'c) 6.01 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 9.73 x 10-6 Ci/ml Antimony-122 (122Sb) 2.70 days 1.01 x 10-5 Ci/ml Xenon-135 (135Xe) 9.10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 1.22 x 10-5 Ci/ml Silver -1 lOm (110mAg) 248.9 days 1.10 x 10-5 Ci/ml 'Listed values are typical of the measured concentrations that exist in the pool coolant at 10 MW, 2 to three days after reactor startup. Due to the structural components that will be used in the target rods and assemblies, it is anticipated that any additional activity added to the pool water will be similar in nature to those found in the table above Page 130 of 190 ATTACHMENT 1 Overall, however, it is anticipated that any increase in activity will not be greater than

as determined by ratio of the TEP thermal power to the MURR core thermal power. It should be noted that the pool water will still be diverted to a pool clean up system (as is currently in place) that has anion-cation removal capability in order to ensure that the pool water is relatively free of contamination. This slight increase in pool water activity is not expected to warrant additional engineering or administrative controls beyond the control currently in place at the MURR. For the purposes of the TEP, cooling water will flow from the WCM, located outside of the reactor pool, through piping located in the pool and then enter at the top of the TA.* After traveling down through the housing the water will then pass upwards through the target rod region where it removes the fission heat from the target rods. The water will then be discharged out of the TA into the pool near the top of the assembly through a diffuser back into the bulk pool water region. Here the water will intermingle with the reactor pool water and circulate through the pool hold-up tank system prior to entering the pool coolant heat exchangers as noted earlier. A portion of that pool water will be diverted off to the TEP cooling system and then cooled further before being reintroduced into the TA. By utilizing existing reactor-related cooling and decay systems for the reduction of N-16, the addition o,f any N-16 produced by the TAs will not add any appreciable source term to the overall inventory. N-16 that must be decayed and/or shielded in order to protect reactor staff. 8.3 Solid Sources Solid -Solid sources are a bit more diverse, but for the most part are very typical of a research reactor facility. Such sources include the reactor fuel in use in the core, irradiated fuel stored in the reactor pool, and new, unirradiated fuel. Additionally, fueled target rods used to produce the Mo-99 will be an additional solid radiation source: After irradiation, the target rods will be to a dedicated hot cell system and then processed to c extract Mo-99 through the SGE process. Any sources of radioactivity germane to the SGE process would be further described in the Part 2 licensing action and not included herein. Thus, the target rods (irradiated and unirradiated) would solely add to the current uranium inventory at MURR. The solid radioactive sources associated with normal operation of MURR are summarized in Table 35. Because the actual inventory of reactor fuel, SGE target rods and other radioactive sources continuously change as part of the normal operation of the reactor and the experimental program, the information presented in Table 35 should be considered representative rather than an exact inventory. Disposition of the irradiated LEU-target material will be discussed in the Part 2 License Amendment submission and is expected to be coordinated through a Uranium Lease Take-Back agreement between MURR and the U.S. Department of Energy (DOE). Page 131of190 ATTACHMENT 1 Table 35 Representative Radioactive Sources at MURR *. '< .: .* .. . . Appt<)ximate Total Sol(rce .. . Nominal Physfoal** Wt%.* *(grams) Description Radioirnclid'e( s) * * . ** Uranium . . . .. *(Ci) . . . U-235 . :. 8 MURR Fuel Highly Enriched NIA In 10 MW Core -93* 6,200 6,656 Elements Uranium Irradiated 45 MURR Fuel Highly Enriched NIA In Pool Storage -93 34,875 37,440 Elements Uranium Irradiated 4MURRFuel Highly Enriched NIA In Storage -New 93 3,100 3,328 Elements Uranium .SGE Target Low-Enriched NIA In Irradiation ---Rods Uranium (LEU) Position 8.4 Radioactive Waste Management Program MURR has a comprehensive Radioactive Waste Management Program that supports operation of the reactor, its ancillary facilities, and their utilization programs. All radioactive waste materials released from the facility through the Ventilation and Air Treatment System, the Radioactive Liquid Waste Retention and Disposal System, a11d the Solid Radioactive Waste Program are identified, assessed, and released or disposed of in conformance with all applicable regulations and in a manner that protects the health and safety of the general public and the environment. It is the policy of the reactor facility to keep the volume of waste materials being generated to the absolute minimum by the efficient use of experiment materials, by the use of proper techniques, and by any other means available. It is not anticipated that additional radiation safety training programs will need to be developed for this portion of the License Amendment request for the irradiation of the LEU target rods. Operations personnel that will be responsible for and perform the actual handling of the individual target rods and assemblies are well versed in the underwater manipulation of materials as demonstrated by the years of successful handling of fuel elements essential to the operation of MURR over the past 50 years of operational experience. At this time, minimal additional Health Physics (HP) procedures are anticipated to need development for the irradiation phase of the SGE target rods. Current reviewed and approved HP procedures related to the operation of the reactor are deemed to be sufficient to protect both the staff and general public during the irradiation phase of this project as it is not significantly different than the actual current operation of the reactor. Any permanently installed radiation monitoring equipment at the reactor facility specifically installed to support the SGE TEF is discussed in detail in Section 4. Any HP-related radiation monitoring equipment Page 132of190 ATTACHMENT 1 used at the MURR is much the same as that described in the MURR SAR. Little additional equipment is anticipated to be needed in order to support the addition and operation of the SGE TEF. Additionally, no changes are anticipated for the MURR HP monitoring and survey program as a result of the irradiation of the target rods nor are any significant number* of additional records anticipated to be generated as a result of this activity. Shielding is the paramount design feature used in controlling radiation exposure during operation of the reactor. Shielding has been installed to keep radiation levels in areas occupied by all personnel ALARA. All shielding thicknesses are based on an operating power level of 10 MW. Fuel storage and handling requirements are based on 40-day continuous operation at 10 MW prior to shutting down and removing fuel. With over 50 years of operational history, the installed radiation shielding has performed more than adequately as designed and analyzed. This same shielding will be used to provide radiation exposure protection from the two (2) target assemblies that will be operating in the graphite reflector region directly outside the reactor pressure vessel and beryllium reflector region. It should be noted however that one entire TEF assembly ( used for the production of Mo-99 via the SGE process contains approximately

grams of U-235; approximately

of the U-235 contained in one single fresh MURR fuel element. Thus, two (2) full target assemblies with

target rods each would contain approximately .. of the U-235 that is contained in the reactor core with -fresh fuel elements. This does not count the fuel elements in the reactor pool that are being cooled for either reuse within the reactor or awaiting sufficient cooling and decay for shipment as spent fuel. Thus the addition of the *

target assemblies adds less than -to the entire source tenn that is contained in the pool due to the use and storage of fuel elements located there. Similar calculations were performed using the computer program MicroShield to predict the dose resulting from the movement of a target cartridge to a position underneath the pool water surface that allows handling and removal of the individual target rods for eventual placement into the transfer cask. These doses are summarized in Table 36 below and provide the worst-case estimate of exposure rates from target cartridge movement activities . MURR staff is not expected to be present in the fields given in Table 36 below on a routine or regular basis. Target cartridge movement activities are anticipated to happen . It is also expected that each handling activity will be rotated amongst several staff members throughout the year. The maximum expected dose rates are presented in Table 36. Page 133 of 190 ATTACHMENT 1 Table 36 Maximum Expected Dose Rates from Target Cartridge Movement Activities One Hour after EOI Dose Location Surface Exposure Rate I foot Exposure Rate I meter Exposure Rate (mR/hr) (mR/hr) (mR/hr) Reactor Pool 127 100 61 Biological Shield at 18 13 6.5 Handling Station In conclusion, it is not apparent that any additional major changes are required to the MURR RPP due to the addition of the irradiation facilities that will be used for the SGE TEP. 9. Conduct of Operations 9.1 Procedures Standard Operating Procedures (SOP) will be developed that are specific for operation and maintenance of the SGE TEP and also reactor operation with the TEP. The following topics will be incorporated into procedures for TEP systems installation, testing, operation, and maintenance.

TEP component/systems installation and testing

TEP startup plan

Target cartridge loading, unloading, and in-pool transfer

Target rod receiving, inspection, and storage

TCS operation

TEP preventative maintenance and inspection

Surveillance procedures required by the Technical Specifications Table 37 lists the SOPs that are expected to be used for routine operation of the facility. Page 134of190 ATTACHMENT 1 Table 37 Standard Operating Procedures No. Title Contents

Receiving and inspection of equipment

Installation checkout of Water Cooling Module

Piping installation 1 TEF Installation and Checkout

Welding and inspection procedures

Instrumentation and control installation and checkout

Target housing installation

Cold commissioning with filler rods

Safety interlock checkout 2 TEF Commissioning and Startup

Hot commissioning

Hot acceptance procedure

Checklists

Transfer from storage

Cartridge loading 3 Target Rod Loading and

Target assembly loading and unloading Unloading

Cartridge unloading and temporary cooling

Checklists

Material control procedure Target Rod Receiving, Inspection

Unpacking and inspection 4 and Storage

Acceptance criteria

Storage procedure

Pre-operational checkout

Startup procedure 5 TEF Normal Operations

Monitors and data recording

Safety interlocks and alanns

Shutdown and standby

Routine maintenance and inspection

Planned equipment replacement 6 TEF Maintenance and Inspection

Equipment performance monitoring

Instrumentation checkout and calibration 9.2 Target Experimental Facility Commissioning Plan A high-level commissioning plan is presented in this section. A detailed startup and commissioning plan will be prepared and approved by the Reactor Safety Procedure Review Subcommittee (a subcommittee of the Reactor Advisory Committee) prior to startup of the TEF. Appropriate hold points will be identified in the plan for review of data and for approval to proceed to the next step of the startup plan Page 135 of 190 ATTACHMENT 1 9.2.1 Assumptions

Secondary Coolant System modifications complete

TCS installation complete outside of the reactor pool

Interface with the MURR reactor safety system ready to be connected

Experimental Facilities reaay to be installed

Both filler and target rods on hand 9.2.2 Phase 1: Installation and Initial Testing with One TEF Target Assembly

Reactor shutdown

Reactor startup with reference core configuration to 50 kW (log ECP control blade heights)

Reactor shutdown

Remove current graphite reflector element from position -* Install one new experimental facility target housing o Connect forced cooling loop o Connect interface to reactor safety system o Install cartridge with. filler rods loaded o Complete coolant system testing including all reactor safety system Technical Specification surveillances

After all cooling testing is complete, remove cartridge with

filler rods

Reactor startup to 50 kW o Calculate reactivity worth with new facility experimental facility target housing

Reactor shutdown

Install cartridge with

filler rods

Make coolant system adjustments due to flow restriction changes

Reactor startup to 50 kW o Calculate reactivity worth of cartridge and

filler rods

Continue reactor startup to 10 MW with forced flow and

filler rods 9.2;3 Phase 2: Loading of One Cartridge

Reactor shutdown Page 136 of 190 ATTACHMENT 1

Reactor startup with reference configuration to 50 kW with all filler target rods

Reactor shutdown

Reactor startup with -to 10 kW with Check thermocouple temperature and target thermal calculations. Verify excess reactivity, shutdown margin, and secured experiment reactivity.

Increase power to 50 kW -check thermocouple temperature and target thermal calculations

Increase power to 100 kW -check thermocouple temperature and target thermal calculations

Increase power to 250 kW -check thermocouple temperature and target thermal calculations

Increase power to 500 kW -check thermocouple temperature and target thermal calculations

Increase power to 1 MW -check thermocouple temperature and target thermal calculations

Increase power to 2 MW -check thermocouple temperature and target thermal calculations

Increase power to 5 MW -check thermocouple temperature and target thermal calculations

Increase power to 10 MW -check thermocouple temperature and target thermal calculations

Operate at 10 MW all week

Visually inspect each target rod at the end of the week after removal 9.2.4 Phase 3: Loading of One Cartridge

Reactor shutdown

Reactor startup with reference configuration to 50 kW with all filler target rods

Reactor shutdown

Reactor startup with

target rods to 10 kW with thermocouple. Check thermocouple temperature and target thermal calculations. Verify excess reactivity, shutdown margin, and secured experiment reactivity. Vary coolant inlet temperature between max and min temperatures to verify effect on reactivity.

Increase power to 50 kW -check thermocouple temperature and target thermal calculations

Increase power to 100 kW -check thermocouple temperature and target thermal calculations

Increase power to 250 kW -check thermocouple temperature and target thermal calculations

Increase power to 500 kW -check thennocouple temperature and target thermal calculations

Increase power to 1 MW -check thermocouple teJ,nperature and target thermal calculations

Increase power to 2 MW -check thermocouple temperature and target thermal calculations

Increase power to 5 MW -check thermocouple temperature and target thermal calculations Page 137 of 190 ATTACHMENT 1

Increase power to 10 MW -check thennocouple temperature and target thennal calculations

Operate at 10 MW all week with

target rods

Visually inspect each target rod at the end of the week after removal 9.2.5 Phase 4: Installation of Second TEF Target Assembly and Initial Testing and Conduct Full .Target Rod Experiment Run

Reactor shutdown

Reactor startup with reference configuration to 50 kW

Reactor shutdown

Remove current graphite reflector element

Install -experimental facility with filler target rods o Connect forced cooling loop o Connect interface to reactor safety system o Install cartridge with

filler rods loaded o Complete coolant system testing including all reactor safety system Technical Specification surveillances

After all cooling testing is complete, remove cartridge with

filler rods

Install both cartridges with

target rods

Reactor startup to 10 kW with thennocouples. Check thermocouples temperature and target thermal calculations. Verify excess reactivity, shutdown margin, and secured experiment reactivity.

Increase power to 50 kW -check thermocouple temperature and target thermal calculations

Increase power to 100 kW -check thermocouple temperature and target thennal calculations

Increase power to 250 kW -check thermocouple temperature and target thermal calculations

Increase power to 500 kW -check thermocouple temperature and target thermal calculations

Increase power to 1 MW -check thennocouple temperature and target thermal calculations

Increase power to 2 MW -check thermocouple temperature and target thermal calculations

Increase power to 5 MW -check thermocouple temperature and target thermal calculations

Increase power to 10 MW -check thermocouple temperature and target thermal calculations

Operate at 10 MW all week with. target rods

Visually inspect each target rod at the end of the week after removal Page 138of190 ATTACHMENT 1 9.3 Material Control & Accounting The LEU target rods will be controlled and accounted for under MURR's existing material control and accounting procedures. The fresh target rods will be stored in the same storage area as reactor fuel. Irradiated target rod material control will be discussed in the Part 2 License Amendment application. 10. Target Experimental Facility Accident Analyses Accidents affecting the SGE TEF while it is in the reactor pool can occur either during reactor operation or after reactor shutdown when the irradiated target rods are being transferred in the reactor pool. Accidents can be initiated by a failure either in the TEF or in the MURR reactor facility. Potential accidents initiated by the TEF include breach of target rod cladding, loss of target coolant from a pipe break, loss of target cooling flow, or mishandling of the target. Accidents initiated by the reactor facility include insertion of excess reactivity, loss of primary coolant flow, loss of primary coolant, loss of pool coolant or loss of offsite electrical power. The breach of target rod cladding has been identified as the TEF maximum hypothetical accident (MHA). 10.1 Target Experimental Facility Maximum Hypothetical Accident An accident resulting in a significant release of fission products to the environment from operating the SGE TEF is considered highly improbable. The redundant safety measures, strictly controlled quality and administrative procedures in the design, fabrication and operation of the facility give high confidence in the improbable nature of such a release. Nevertheless, inadvertent physical damage to the TEF from manufacturing defects or damage from handling must be considered and analyzed for licensing purposes with a defined, enveloping event that can be used for bounding the radiation hazard. For the TEF, such an enveloping event, the MHA, is a postulated breach of a target rod cladding during full power operation. The initiating event for this is hypothesized to be caused by a defect in the cladding or a defect in either of the welds that attach the upper and lower end fittings. It is further postulated that this accident occurs at the worst possible time at the end of an irradiation which maximizes the fission product inventory in the gas gap between the target material and the cladding. 10.1.1 Description of Maximum Hypothetical Accident Scenario The scenario for the MHA is a postulated breach from a defective weld or cladding at the end of --irradiation. The fission gas in the void volume of the target rod is released to the target cooling water which is then discharged to the reactor pool water, then into the reactor containment building, and ultimately a release to the outside environment. During operation, fission of the LEU

target material produces a variety of fission products including the fission gases krypton, xenon, and iodine. These fission gases can diffuse out of the

target material and into the void spaces within the target rod. As the target rod bums its U-235, there is an accumulation of fission gases in these void spaces. A breach of a target rod at the end of its -design life will release this accumulated fission gas inventory. Originally, the -pressure inside the Page 139of190 ATTACHMENT 1 . Temperature, thermal expansion and fission gas release raise the pressure to Therefore, a breach of the target rod rapidly releases the fission gap gas into the target cooling water which is then discharged to the reactor pool immediately above the target rods. 10.1.2 Source Term The radionuclide inventory in the target rods at the end of irradiation was calculated by using MCODE, a code to couple MCNP6 and ORIGEN2, to predict isotopic concentration at the end of irradiation. A Once Through Source Term Model (OTSTM) was developed to predict the evolution of radionuclides after irradiation (Reference 35). This program uses an analytic solution to a generalized Bateman equation to calculate the nuclide inventory as a function of time. In addition, OTSTM calculates the decay heat from the actinides and fission products according to ANSI/ ANS-5.1-2014, "Decay Heat Power in Light Water Reactors" (Reference 36), and the fraction of the volatile nuclides that may be released from a target rod gap ifthe cladding is breached according to ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel" (Reference 37). The subsequent decay of 1307 nuclides as they move through the TEF was also modeled using OTSTM. In the OTSTM fission products diffuse out of the

matrix and into the pellet-cladding gap during irradiation. This "gap gas" can then escape into the atmosphere if the cladding is breached during an accident and is the bases for the MHA source tenn. The source term for the gap gas is calculated using the factor of 5X conservatism on the release to birth ratios, as recommended in ANSI/ANS-5.4-2011 (Reference 37) and to be conservative, the source tenn for the gap gas is calculated using power density and temperature profiles in the target rod at the beginning and the burnup at the end of a three-week irradiation. The temperature is highest at the beginning because the power density is highest in the fresh pellets and the burnup is highest at the end, giving the highest pellet surface to volume ratio. In addition, the release fraction for the hottest, highest power target rod is used for calculating the gap gas for all the rods in a set. The total activity of all isotopes of iodine, krypton and xenon in the gap gas of the hottest target rod is

Ci at 0 hours0 days 0 hours 0 weeks 0 months (End of Irradiation). Note that the sum of the activities of the individual nuclides in Table 38 below is slightly less than the total activity because Table 38 does not include some short lived nuclides with minor radiological consequence. The activity in a

target rod set is obtained by multiplying the activity in the hottest rod by .. , which is the ratio of the total fission power in a

rod set to the fission power in the hottest rod. Note that most of the fission products are volatile during irradiation due to the very high temperatures inside the rods. However, we assume that only isotopes of iodine, krypton and xenon are volatile after irradiation or upon target rod cladding failure because the rods are then much cooler. Nonetheless, the decay and progeny of non-volatile nuclides, such as Te-132, does contribute to the activities shown in Table 38. Page 140of190 ATTACHMENT 1 More detailed information about the source term calculation, including the input and output files for the codes used, can be found in GA Report 30441R00022, "Source Tenn Analysis Design Calculation Report" (Reference 38). However, for ease of reviewability the source tenn output applicable to the MHA is repeated in Table 38. Table 38 Activity of Volatile Fission Products in the Gap Gas of the Hottest Target Rod Fission Product Half-life Activity at EOI Activity _at 6' hours (ti) (Ci) .. Kr-85m 4.481 h --Kr-85 10.72 y --Kr-87 1.272 h --Kr-88 2.839 h --Kr-89 3.15 m --Kr-90 32.32 s --I-131 8.041 d - 132 2.3 h - 133 20.8 h - 134 52.6m - 135 6.611 h --Xe-133 5.245 d --Xe-135 9.089 h --Xe-135m 15.29 m --Xe-137 3.818 m --Xe-138 14.08 m --Xe-139 39.68 s --Totals: --10.1.3 Dose Assessment The methodology applied for the dose assessment calculations presented throughout this section are based on the fueled experiment failure methodology submitted by MURR to the NRC as part of relicensing efforts. 10.1.4 Radionuclide Concentration in Reactor Pool Water A breach of the target rod with the highest power generation at the end of a -irradiation cycle will release the iodine, krypton and xenon gap activities shown in Table 39 into the MURR reactor pool. Page 141of190 ATTACHMENT 1 Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors. However, for the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5 minute evacuation period is used as the basis for the stay time in the dose calculations for personnel that are in containment during target failure. Table 39 Iodine and Noble Gas Activities Released to MURR Reactor Pool Activity ( Ci) 1311 -2.6 3 x 10+07 85Kr-3.16 x 10+04 133Xe -6.48 x 10+07 1321 -1.4 5 x 10+08 85mKr -5.96 X 10+06 135Xe -8.11 x 10+06 1331 -4.5 8 x 10+07 87Kr-6.29 x 10+06 135mxe -5.51 x 10+06 1341 -2.6 1 x 10+07 88Kr-1.21 x 10+07 137Xe -3.05 x 10+06 1351 -3.0 6 x 10+07 89Kr -2.27 x 10+06 138Xe -5.91 x 10+06 90Kr -8.91 x 10+05 139Xe -1.04 x 1 o+06 The iodine released into the reactor pool over a 5 minute period is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which then results in the following pool water concentrations for the iodine isotopes (Table 40). The water solubility of the krypton and xenon noble gases released into the pool over this same time period is conservatively ignored. The gas bubble rise time in the reactor pool from where the TEP is situated in the graphite reflector region to the pool surface has been measured at 17 seconds, so it is assumed the earliest any of the radioisotopes from the TEP, including radioiodine nuclides, enters into the containment building air volume is after 17 seconds. It is also assumed when the radioactivity enters the containment air volume it instantaneously forms a uniform concentration in the isolated containment structure. Table 40 Iodine Concentrations in Pool Water Concentration (µCi/2al) 1311 -1.31 x 10+03 1321 -7.23 x 10+03 1331 -2.29 x 1 o+03 1341 -1.30 x 10+03 1351 -1.53 x 10+03 Page 142of190 ATTACHMENT 1 10.1.5 Radionuclide Concentration in Containment When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed iodine concentrations evaporates into the containment building over the 5 minute period. Containment air with a temperature of 75 °F (23.9 °C) and 100% relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of20 gallons (75.7 L) of water vapor will not cause the containment air to become supersaturated. It is also conservatively assumed that all of the iodine activity in the 20 gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following iodine concentrations in the 225,000 ft3 (6,371.3 m3) containment air volume: Example calculation of the average 1311 concentration released into the containment air during the first minute: 1311 concentration in pool water/gal x 20 gal/5 minx 0.5 minx EXP(-0.693 x (17+30 sec) I (8.02 day x 8.64 x 10+04 sec/day)) I (225,000 ft3 x 28,317 ml/ft3) 1.31 x 10+03 µCi/gal x 2 gal x 0.99995 I (6.371 x 10+09 ml) 4.13 x 1 o-07 µCi/ml The same calculation is used for the other iodine isotopes listed below and was performed for each I-minute interval in the 5 minute period. The average radioiodine concentrations over the 5 minute evacuation period is the average of the five I-minute intervals in the 5 minute period (Table 41). Table 41 Average Iodine Concentrations in the Containment Building Air During the 5 Minute Evacuation Period Concentration (µCi/cc) 1311 -2.06 x 10-06 1321 -1.12 x 10-05 1331 -3.59 x 10-06 1341 -1.95 x 1 o-06 1351 -2.38 x 10-06 As noted previously, the krypton and xenon gases released into the reactor pool from the SGE TEF during the 5 minute evacuation period following a cladding failure are assumed to have no absorption in the pool Page 143 of 190 ATTACHMENT 1 water, rise through the pool in 17 seconds (thus slightly decaying) and enter the containment building air volume where they are assumed to instantaneously fonn a unifonn concentration in the isolated structure. Based on the 225,000-ft3 volume of contaimnent building air, and the previously listed curie quantities of these gases released into the reactor pool, the average noble gas concentrations in the containment structure over the 5 minute evacuation period would be calculated as follows: Example calculation of the average 85Kr concentration in containment during the first minute after the gas enters the containment air: 85Kr activity x EXP(-0.693 x (17+ 30 sec) I (10.76 yrs x 3.156 x 10+07 sec/yr)) I (225,000 ft3 x 28,317 ml/ft3) 3.16 x 10+04 µCi x 0.99999 I (6.371 x 10+09 ml) 4.96 x 1 o-06 µCi/ml The same calculation is used for the other krypton and xenon isotopes listed below in Table 42 and was performed for each I-minute interval in the 5 minute period. The average concentrations over the five minute evacuation period are the average of the five I-minute intervals in the 5 minute period. Table 42 Average Noble Gas Concentrations in the Containment Building Air during the 5 Minute Evacuation Period Concentration (µCi/cc) 85Kr -4.96 x I o-06 133Xe -1.02 x 10-02 85mKr -9.29 x 10-04 135Xe -1.27 x 10-03 87Kr-9.63 x 10-04 135mxe -7.64 x 10-04 88Kr-1.87 x 10-03 137Xe -2.99 x 10-04 89Kr-2.02 x 10-04 138Xe -8.11 x 10-04 90Kr-1.41 x 10-05 139Xe -2.19 x 10-05 10.1.6 Dose Assessment in Restricted Area The objective of this calculation is to present a worst-case occupational dose assessment for an individual who remains in the containment building for 5 minutes following the MHA. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and unifonnly distributed into the building once released into the air. Page 144of190 ATTACHMENT 1 Based on the source tenn data provided, it is possible to detennine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building. Because the airborne radioiodine source is composed of five different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and to then sum the results. Dose multiplication factors were established using the Derived Air Concentrations (DACs) for the "listed" isotopes in Appendix B of 10 CFR 20 and calculated values for the "unlisted" submersion isotopes (Kr-89, Kr-90, Xe-137, and Xe-139). The submersion DAC values that were calculated were done in accordance with the data and methodology as supplied in Federal Guidance Report No. 12. Example calculation of thyroid dose due to rnl: The DAC can also be defined as 50,000 mrem (thyroid target organ limit) I 2,000 hr, or 25 hr. Additionally, 5 minutes of one DAC-hr is 8.33 x 10*02 DAC-hr. 1311 concentration in containment 2.06 x 10*06 µCi/ml rnl DAC (10 CFR 20) 2.00 x 10*08 µCi/ml Dose Multiplication Factor (1311 concentration) I (1311 DAC) (2.06 x 10*06 µCi/ml) I (2.00 x 10*08 µCi/ml) 1.03 x 10+02 Therefore, the 5-minute thyroid exposure from 1311 is: Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr 1.03 x 10+02 x (25 mrem/DAC-hr) x (8.33 x 10*02 DAC-hr) 2.15 x 10+02 1mem The same calculation sequence was repeated for all iodine isotopes and is summarized in Table 43. Page 145of190 ATTACHMENT 1 Table 43 Derived Air Concentration Values and 5-Minute Exposures for Iodine Radionuclide Derived Air Concentration 5-Minute Exposure (µCi/ml) (mrem-CDE) 1311 2.00 x 10-08 2.15 x 10+02 1321 3.00 x 10-06 7.74 x lO+OO 1331 1.00 x 10-07 7.47 x lO+OI 1341 2.00 x 10-05 2.03 X 10-0I 1351 7.00 x 10-07 7.09x lO+oo Doses from the krypton and xenon radionuclides present in the containment building are assessed in much the same manner as the iodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes in containment were based on their average concentration in the containment air and the corresponding DAC. The whole body dose due to 85Kr is calculated as follows: The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mrem/DAC-hr. Additionally, 5 minutes of one DAC-hr is 8.33 x 10-02 DAC-hr. 85Kr concentration in containment 85Kr DAC (10 CFR 20) = 4.96 x 10-06 µCi/ml 1.00 x 1 o-04 µCi/ml Dose Multiplication Factor (85Kr concentration) I (85Kr DAC) (4.96 x 10-06 µCi/ml) I (l.00 x 10-04 µCi/ml) = 4.96 x 10-02 Therefore, a 5 minute whole body exposure from 85Kr is: = Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr = 4.96 x 10-02 x (2.5 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr) = 1.03 x 10-02 mrem The same calculation sequence was repeated for all noble gases and is summarized in Table 44. Page 146of190 ATTACHMENT 1 Table 44 Derived Air Concentration Values and 5 Minute Exposures -Noble Gases Radionuclide Derived Air Concentration 5-Minute Exposure (µCi/ml) (mrem-CEDE) 85Kr 1.00 x 10-04 1.03 x 10-02 85mKr 2.00 x 10-05 9.68 x 10+-00 87Kr 5.00 x 10-06 4.01 x 1o+OI 88Kr 2.00 x 10-06 1.95 x 10+02 89Kr 1.90 x 10-06 2.22 x 1o+OI 90Kr 2.80 x 10-06 1.05 X 10+-00 133Xe 1.00 x 10-04 2.12 x lO+OI 135Xe 1.00 x 10-05 2.64 x 1o+OI 135mxe 9.00 x 10-06 1.77 x 10+-01 137Xe 2.00 x 10-05 3.11x10+00 138Xe 4.00 x 10-06 4.22 x 1o+OI 139Xe 3.70 x 10-06 1.24 x 10+00 To finalize the occupational dose in tenns of Total Effective Dose Equivalent (TEDE) for a 5 minute exposure in the containment building during the MHA, the doses from the iodines and noble gases must be added together. This summation results in the following TEDE values presented in Table 45. Table 45 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building Iodine -Committed Dose Equivalent (Thyroid) 304 mrem Iodine -Committed Effective Dose Equivalent (CDE x 0.03) 9mrem Noble Gas -Committed Effective Dose Equivalent 380 mrem Total Effective Dose Equivalent 389 mrem By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for those exposed during a TEP failure and the MHA to applicable NRC dose limits in 10 CPR 20, the final values are shown to be well within the published regulatory limits, in fact, less than 10% of any occupational

limit. Page 147 of 190 ATTACHMENT 1 10.1. 7 Dose Consequences to Members of the Public As noted earlier in this chapter, the containment building ventilation system will shut down and the containment building will be isolated from the surrounding areas upon actuation of the isolation system. A breach of target rod cladding will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere. It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure change has occurred in conjunction with the target failure. A reasonable assumption would be a pressure change on the order of 0. 7 inches of Hg (25.4 mm of Hg at 22 °C), which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment structure. Making the conservative assumption that the containment building will leak at the Technical Specification leakage rate limit, 10% of the contained volume over a 24-hour period from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the air leakage from the containment structure in standard cubic feet per minute (scfm) as a function of containment pressure can be expressed by the following equation: LR 17.68 x (CP-14.7)112; where: LR leakage rate from containment (scfm); and CP containment pressure (psia). The minimum Technical Specification free volume of the containment building is 225,000-ft3 at standard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the containment structure would hold approximately 255,612 standard cubic feet (scf) of air. When applying the Technical Specification leakage rate equation to the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for the leak rate to decrease to zero from an initial leakage rate of approximately 10.25 scfm, which would occur at the start of the event. Determination of the average leakage rate is subdivided into multiple intervals within the total 16.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months leak duration to provide a more accurate calculation of release concentrations from the facility using the Technical Specification leakage rate equation. The average containment building leakage rate for each of the first five (5) 1-hour intervals and for the following three (3) 4-hour intervals is provided below in Table 46. Page 148of190 ATTACHMENT 1 Table 46 Average Containment Building Leakage Rate Hours: 0-1 1-2 2-3 3-4 4-5 5-9 9-12 12-16.5 scf/hr: 595.6 558.7 521.8 485.0 448.1 355.8 207.9 67.9 Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following target failure. First of all, most leakage pathways from contaimnent discharge into the reactor laboratory building, which surrounds the containment structure. Since the laboratory building ventilation system continues to operate during target failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 cfm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in a contaimnent mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75% of the iodine released will be deposited in the containment vessel. For the purposes of this analysis a conservative value from Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," was used and provides a reduction factor of 50% for plate-out and deposition of iodine within the containment building. Thus, due to this 50% iodine deposition in the contaimnent building, each cubic foot of air released from contaimnent has a iodine concentration that is fifty percent of each cubic foot within the containment building air. Example calculation of average 1311 concentration in contaimnent during the first hour: (1311 concentration in pool water/gal x 20 gal) x EXP[(-0.693 x 0.5 hr) I (1311 T112)] I (scf in containment) (1.31 x 10+o3 µCi/gal x 20 gal) x 0.99820 I (229,800 scf) 1.14 x 10*01 µCi/scf The average 1311 concentration in containment over the first hour is then further reduced by the 50% iodine plate-out reduction factor prior to escaping from containment at the average leak rate and further diluted in 30,500 scfm of laboratory building exhaust ventilation prior to being released from the facility. Page 149of190 ATTACHMENT 1 Example calculation of average 1311 concentration in air released from the facility during the first hour: (average 1311 concentration in containment µCi/scf x average leak rate x 0.5 reduction factor) I (30,500 scfm x 60 min/hr x 28,317 ml/scf) (1.14 x 10-01µCi/scfx595.6 x 0.5)/(5.18 x 10+10) 6.56 x 10-10 µCi/ml The average of first five (5) 1-hour intervals and the following three (3) 4-hour intervals is then summed and averaged over the entire 16.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months duration to calculate an average of the air concentration released from the facility. The same calculation is used for the other radioiodines and is provided in Table 47. Table 47 Average Iodine Concentrations in Air Exiting the Exhaust Stack Activities (µCi/ml) 1311 -3.38 x 10-10 1321 -5.89 X 10-lO 1331 -5.01 X 10-JO 1341 -4.60 x 10-ll 1351 -2.39 x 10-10 Calculation of nobles gases released through the exhaust stack is identical to the calculation of the radioiodines released except no credit is taken for activity absorbed by pool water or reduction via out inside of the containment building. Example calculation of average 85Kr concentration in containment during the first hour: 85Kr activity x EXP[(-0.693 x 0.5 hr) I (85Kr T112)] I (scf in containment) 3.16 x 10+04 µCi x 0.99999 I (229,800 scf) 1.38 x 10-01 µCi/scf Example calculation of average 85Kr concentration in air released from the facility during the first hour: (average 85Kr concentration in containment µCi/scf x average leak rate) I (30,500 scfm x 60 min/hr x 28,317 ml/scf) (1.38 x 10-01 µCi/scf x 595.6) I (5.18 x 10+10) 1.58 x 10-09 µCi/ml Page 150of190 ATTACHMENT 1 The average of the first five (5) I-hour intervals and the following three (3) 4-hour intervals is then summed and averaged over the entire 16.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months duration to calculate an average of the air concentration released from the facility. The same calculation is used for the other noble gasses and is provided in Table 48. Table 48 Noble Gas Concentrations in Air Exiting the Exhaust Stack Activities (µCi/ml) 85Kr-8.29 X 10-IO 133Xe -1.65 x 10-06 85mKr -7.67 X 10-08 135Xe -1.44 x 10-07 87Kr-3.13 x 10-08 135mxe -4.57 x 10-09 88Kr -1.16 x 10-01 137Xe -4.00 X 10-ll 89Kr -9.34 x 10-12 138Xe -4.31 x 1 o-09 90Kr-4.56 x 10-26 139Xe -7.10 x 10-23 10.1.8 Dose Assessment in Unrestricted Area Assuming that (1) the above provided average leak rates from the reactor containment building, (2) the leak continues for about 16.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292 (explained below) and (5) there is no decay of any iodines or noble gases, then the following concentrations of iodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are at the point of maximum concentration in the unrestricted area assuming uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of contaimnent leakage following target failure. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months that the containment building was leaking. A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis. It is assumed that all offsite (public) doses occur under these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. This point conservatively assumes a Stability Class F; which normally occurs only 11.4% of the time when the wind blows from the south. Thus, this calculation is conservative. Page 151 of 190 ATTACHMENT 1 10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes. Effluent Concentration Limits were calculated for each of the four ( 4) "unlisted" noble gases (Kr-89, Kr-90, Xe-137 and Xe-138) using the data and methodology contained in Federal Guidance Report No. 12 for submersion isotopes. The DAC value was first calculated and then a factor of 219 was applied using 10 CFR 20, Appendix B, methodology for effluent values from submersion isotopes. Exposure at 1 DAC equates to 5000 mrem per year whereas the Effluent Concentration Limit is 50 mrem per year. Thus, there is a factor of 100 times lower allowable dose for the Effluent Concentration Limit as compared to the DAC. Exposure at the effluent concentration limit assumes a person is in that effluent concentration for 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year. Thus, the time assumed for exposure to the effluent concentration limit is a factor of 4.38 longer than the 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months per year that defines a DAC. No credit is taken for transit time from the stack to the receptor point. In the case of Kr-89 and Xe-137 the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) lives. Example calculation of whole body dose in the unrestricted area due to 1311: Conversion Factor: (Public dose limit of 50 mreni/yr) x (1 yr/8760 hours)= 5.71 x 10-03 mrem/hr 1311 concentration 1311 effluent concentration limit 1311 Conversion Factor 3.38 x 10-lO µCi/ml 2.00 x 10-10 µCi/ml 5.71 x 10-03 mrem/hr Therefore, a 16.5-hour whole body exposure from 1311 at the maximum receptor site is: = (1311 concentration I 292 dilution factor) I (1311 effluent concentration limit) x (Conversion Factor x 16.5 hrs) (3.38 x 10-10 µCi/ml I 292) I (2.00 x 10-10 µCi/ml) x (5.71 x 10-03 mrem/hr x 16.5 hrs) 5.45 x 10-04 mrem The same calculation is used for the other isotopes (iodines and noble gases) and results are listed in Table 49 and Table 50. Page 152of190 ATTACHMENT 1 Table 49 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Iodine Effluent Limit

Concentration1 at Radiation Dose Radioiodine (µCi/ml) Maximum Receptor Site (mrem) (µCi/ml) 1311 2.00 x 10-10 1.16 x 10-12 5.45 x 10-04 1321 2.00 x 10-08 2.02 x 10-12 9.51 x 10-0G 1331 1.00 x 10-09 1.72 x 10-12 1.62 x 10-04 1341 6.00 x 10-08 1.58 x 10-13 2.48 x 10-07 1351 6.00 x 10-09 8.18 x 10-13 1.28 x 10-05 Total = 0.000729 Note 1: Concentrations are the average radio iodine concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292. Table 50 Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases Effluent Limit Concentration1 at Radiation Dose Radioisotope (µCi/ml) Maximum Receptor Site (mrem) (µCi/ml) 85Kr 7.00 x 10-01 2.84 x 10*12 3.82 x 10-07 85mKr 1.00 x 10*07 2.63 x 10-10 2.48 x 10-04 87Kr 2.00 x 10-08 1.07 x 10-10 5.06 x 10-04 88Kr 9.00 x 10-09 3.97 x 10*10 4.15 x 10-03 89Kr 2.00 x 10-08 3.20 x 10-14 3.51 x 10-07 90Kr 2.00 x 10-03 1.56 x 10-28 1.23 x 10-21 133Xe 5.00 x 10-07 5.64 x 10*09 1.06 x 10*03 135Xe 7.00 x 10-03 4.94 x 10*10 6.64 x 10-04 135mxe 4.00 x 10-08 1.57 x 10-11 3.69 x 10-05 137Xe 2.00 x 10-08 1.37 x 10*13 1.42 x 10-07 138Xe 2.00 x 10-08 1.48 x 10-11 6.96 x 10-05 139Xe 2.00 x 10-08 2.43 x 10*25 1.43 x 10*13 Total= 0.00674 Note 1: Concentrations are the average noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. Page 153of190 ATTACHMENT 1 To finalize the unrestricted dose in terms of TEDE, the doses from the iodines and noble gases must be added together. Results are provided in Table 51. Table 51 Dose from Iodines and Noble Gases in the Unrestricted Area Committed Effective Dose Equivalent (Iodine) 0.000729 mrem Committed Effective Dose Equivalent (Noble Gases) 0.00674 mrem Total Effective Dose Equivalent 0.00747 mrem Summing the doses from the noble gases and the iodines simply substantiates earlier statements regarding the low levels in the unrestricted area should a failure of the TEF occur and subsequent containment building following such an event. The overall TEDE is much less than 1 mrem, a value far below the 100 mrem regulatory limit within 10 CFR 20 for the unrestricted area. 10.1.9 Radiation Shine through Containment An evaluation of radiation shine through the contaimnent building has also been perfonned to ensure whole body doses are of minimal concern during an accident scenario. The radiation shine model uses MicroShield Ver. 8.02 (Grove Software) for shielding and exposure rate calculations to evaluate radiation shine through the containment structure under accident conditions, and to determine dose consequences to the public and MURR staff. Calculations of exposure rate from the TEF failure were perfonned using a Rectangular Volume -External Dose Point geometry for the representation of the containment structure. The exposure rate values provided below represents the radiation fields at 1 foot (30 cm) from a 12-inch thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ) boundary of 150 meters (492.1 ft) representing the anticipated exposure rates in the restricted and unrestricted area respectively. The airborne concentrations used in calculating the exposure rate values were derived from the total activities of the gap gas nuclides provided in Table 52 divided by the containment free volume of 225,000 scf. Therefore, all radiation shine exposure rate values provided are very conservative as solubility of the iodine nuclides in the pool water, and decay of all nuclides is ignored. The source term also assumes a homogenous mixture of the nuclides within the containment free volume. Table 52 provides the exposure rates external to the containment building at 1 foot and 150 meters from the building surface. Table 52 Radiation Shine through the Containment Building Exposure Rate at 1-Foot from Containment Building Wall 67 mR/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters) 0.32 mR/hr Page 154of190 ATTACHMENT 1 10.2 Insertion of Excess Reactivity Because of (1) the subcriticality of the target cartridges and (2) the physical impossibility of removing a target cartridge without scramming the reactor, a reactivity accident associated with the removal or insertion of the target cartridges is not considered credible. Removal and insertion of the target cartridges only occurs when the MURR reactor is shutdown. Two (2) different reactivity insertion accidents for the MURR reactor core would have an impact on the TA. First, a step insertion of positive reactivity based upon the maximum step insertion that the MURR core can withstand with no core damage, and second, a continuous ramp insertion of positive reactivity based on the continuous withdrawal ofMURR's four (4) shim control blades. 10.2.1 Rapid Insertion of Positive Reactivity For the Insertion of Excess Reactivity accident analysis, the licensed maximum power level of 10 MW was originally used in the reactor SAR as the starting assumption since MURR does not, nor can it legally, operate above this power level. NUREG-1537, Part 2, page 13-9, Standard Review Plan and Acceptance Criteria, states for Insertion of Excess Reactivity accident, that "The accident scenario assumes that the reactor has a maximum load of fuel (consistent with the technical specifications), the reactor is operating at full licensed power, and the control system ... " The accident was reanalyzed at a much more conservative starting power level (11.5 MW) than is required by NUREG-1537 and the results are provided below. 11.5 MW was chosen, instead of the Limiting Safety System Setting (LSSS) set point of 12.5 MW, because the rod run-in system will initiate a rod run-in at 11.5 MW (Technical Specification 3.2.f.1) and shut down the reactor prior to reaching the LSSS SCRAM set point of 125%. The third paragraph on Page 13-17 of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivity accident analysis. For both the reactor SAR analyses, as well as for the updated analysis presented here, the CB insertion times are based on the current and relicensing Technical Specification 3.2.c requirement of insertion to the 20% withdrawn position in less than 0.7 seconds. The insertion rate was calculated based on the shim CBs travelling from 26 inches (fully withdrawn) to 5.2 inches (20% withdrawn or 80% inserted) in 0.7 seconds. This is a conservative assumption because monthly control blade drop time verifications performed at MURR have always yielded insertion times of 0.6 seconds or less. Similar to the reactor SAR analysis, the Reactivity Transient Analysis program PARET (V7.5), maintained and distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) was used. For the Insertion of Excess Reactivity accident analysis, two (2) channels were modeled in P ARET; a hot channel representing worst-case conditions inside the core and an average channel representing the rest of the core experiencing "average" conditions. As indicated earlier, the transient was started from an initial power level of 11.5 MW with core coolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm and 155 °F, respectively. Also, pressurizer pressure was at 75 psia (LSSS value). Since the Insertion of Excess Reactivity transient was analyzed from a Page 155of190 ATTACHMENT 1 starting power level of 11.5 MW, the rod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only the high power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds was incorporated into the CB scram model so that the CBs would only start to insert 0.15 seconds after the power level had exceeded the scram set point of 12.5 MW. The results of a step reactivity insertion of 600 pcm (+0.006 are shown below in Figure 75. As expected, due to the higher starting core power level, much lower core coolant flow rate and much higher than nonnal core coolant inlet temperature conditions assumed for this updated analysis, the peak power during the transient momentarily reaches approximately 37.4 MW compared to a value of approximately 33.0 MW reported in prior SAR analyses for the same 600 pcm step reactivity insertion. 40.00 35.00 +----------ff------------------+-350 -POWER MW 0.00 0 0.00 0.50 1.00 1.50 2.00 2.50 3.00 Time (seconds) Figure 75 Reactor Power, Fuel and Cladding Temperatures vs. Time for a Positive Reactivity Step Insertion of 0.006 Ak/k The power generation in the T As would follow the same proportional power transient that the reactor core experiences because the TAs are driven by the neutron flux generated by the reactor core. The target rod analysis for the positive reactivity step insertion examines the maximum powered target rod at the beginning and end of a three-week irradiation. A steady-state analysis of the three-week irradiation was perfonned using FRAPCON to establish the initial conditions for the transient analysis. The FRAPCON analysis assumed that the first and last day of the three-week irradiation are at 115% power and the rest of the operating period is at 100% power. The FRAPCON analysis also Page 156of190 ATTACHMENT 1 assumes that the TA flow rate is 85% of nominal during the first and last day of operation. Mid-week and weekend shutdowns are also included in the power history of the TA. The transient analysis was perfonned using the FRAPTRAN code. FRAPCON and FRAPTRAN are two NRC-sponsored computer codes that can model the steady-state and transient thennal-mechanical behavior of light water reactor oxide fuel. Phenomena modeled by the codes include heat transfer through fuel and cladding to coolant, cladding elastic/plastic deformation, fuel-cladding mechanical interaction, fission gas release, rod internal pressure, and cladding oxidation. The FRAPCON steady-state analysis defines the bumup,

and cladding deformation, and fission gas release that fonn the starting point for the FRAPTRAN analysis. The FRAPTRAN analysis increases the power to 115% and decreases the flow rate to

within 40 seconds. Twenty-one seconds later, the 600 pcm (+0.006 step reactivity insertion is simulated. The target cooling water inlet temperature to the TA was assumed to be at which corresponds to an outlet temperature at the maximum pool temperature of during nominal conditions. The maximum cladding strain occurs with the minimum gap tolerance . The maximum cladding strain also occurs at the end of irradiation due to the effects of bum up on gap closure and fission gas release. The FRAPCON and FRAPTRAN analysis also use the maximum tolerance on the -diameter that forms the flow area of the target cartridge. This tolerance results in the minimum coolant velocity consistent with the minimum gap. The analysis also uses the maximum pellet outer diameter, minimum cladding inner diameter, minimum cladding thickness, and minimum cladding outer diameter. The strain transient that the cladding is predicted to undergo during the positive step reactivity insertion is shown in Figure 76. The hoop strain increases from about -to just under -but remains under the 1 % strain limit. Maximum temperatures occur with the maximum gap tolerance of rather than the minimum gap tolerance. Bumup effects reduce the gap so that the maximum pellet temperature occurs at the beginning of irradiation. The analysis also uses the minimum pellet outer diameter, maximum cladding inner diameter, maximum cladding thickness, and maximum cladding outer diameter in order to obtain the maximum gap tolerance or to maximize peak pellet temperature. The peak target temperature transient is shown in Figure 77. The peak target temperature increases from to slightly less than in less than 0.4 seconds. This peak

temperature is below the melting temperature Page 157of190 r-----------------------------------ATTACHMENT 1 Time (seconds) Figure 76 Cladding Strains at Peak Pellet Location During a Positive 0.006 Ak/k Reactivity Insertion Time (seconds) Figure 77 Peak Target Pellet Temperature During a Positive 0.006 Ak/k Reactivity Insertion Page 158of190 ATTACHMENT 1 The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where the peak pellet centerline temperature occurs is shown in Figure 78. The cladding outer diameter temperature shows a modest increase in temperature of around -* The pellet outer diameter and cladding inner diameter temperatures become more closely coupled due to closure of the gap between the pellet and cladding and an increase in interface pressure. Cladding hoop strain for the -gap at the end of irradiation goes through a similar transient as for the -gap but increases from -to just under-Peak heat generation within the target pellets is a factor of. greater than nominal. Heat capacity and thennal resistances result in the surface heat flux at the peak location being only a factor of. to

greater than nominal, depending on the initial gap size and burnup conditions. FRAPTRAN assesses critical heat flux (CHF) using one of five different CHF correlations. The Macbeth CHF correlation was chosen because of its validity at low pressures. The Macbeth CHFR reaches a minimum value of between 2.16 and 2.24 depending on initial gap size and burnup conditions. The Bernath CHFR reaches a minimum of 1.67 to 1.87 depending on the case. No target rod damage or radiological release would occur from this accident. Time (seconds) Figure 78 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a Positive 0.006 Ak/k Reactivity Insertion Page 159of190 ATTACHMENT 1 10.3 Control Blade Withdrawal As presented in Section 13.2.2.1.2 of the MURR SAR, a positive reactivity ramp insertion rate of 0.0003 which is the Technical Specification limit on the maximum rate of reactivity insertion for all four (4) shim CBs operating simultaneously, was introduced to the reactor starting at subcritical cold conditions and at an initial power level of 10 MW. For subcritical cold conditions, the short period reactor scram tenninates the transient within 150 sec before the power has reached 64 watts. No target pellet or cladding damage would occur at such a low power. For full power conditions, the high power reactor scram tenninates the transient after 4.53 seconds when reactor power rises from 10.0 MW to the 12.5 MW high power scram set point. The thermal-mechanical perfonnance of the target rod during this transient was analyzed using FRAPCON and FRAPTRAN. The power transient that drives the heat generation in the target rods is shown in Figure 79. The figure includes 1 sec of steady-state operation at 100% power and the power transient after reactor SCRAM. j '5 c .2 g ..... 0.1 0 2 3 4 5 6 7 Time (seconds) Figure 79 Power Transient During a 0.0003 Ak/k per Second Reactivity Insertion The target rod analysis for the CB withdrawal examines the maximum powered target rod at the beginning and end of -irradiation. A steady-state analysis of the -irradiation was perfonned at I 00% power using FRAPCON to establish the initial conditions for the transient analysis. The FRAPCON analysis also assumes that the TA flow rate is I 00% of nominal during the -irradiation. Mid-week and weekend shutdowns are also included in the power history for the TA. Page 160of190

ATTACHMENT 1 The FRAPCON steady-state analysis defines the bumup,

and cladding deformation, and fission gas release that form the starting point for the FRAPTRAN analysis. The maximum cladding strain occurs with the minimum gap tolerance . The maximum cladding strain also occurs at the end of the -irradiation due to the effects of bumup on gap closure and fission gas release. The strain transient that the cladding is predicted to undergo during the control blade withdrawal is shown in Figure 80. The hoop strain increases from about -to just over -but remains well under the 1 % strain limit. Time (seconds) Figure 80 Cladding Strains at Peak Pellet Location During a 0.0003 Ak/k per Second Reactivity Insertion For the case using the -gap between the pellet and cladding, the pellet centerline temperature increases just over at the beginning of irradiation. Bumup effects reduce the gap so that the maximum pellet temperature decreases during irradiation. The temperature of the cladding inner diameter increases to just greater than temperature increases by less than-* The FRAPTRAN analysis of the , and the cladding outer diameter gap case predicts the maximum pellet temperatures due to the higher thermal resistance between the pellet and cladding. The maximum pellet temperature increases from at the beginning of irradiation as shown in Figure 81. The pellet outer diameter and cladding inner and outer diameter temperature at the axial location where the peak pellet centerline temperature occurs is shown in Figure 82. The pellet outer diameter and cladding inner diameter temperatures become more closely coupled due to closure of the gap between the pellet and cladding and an increase in interface pressure. The cladding inner diameter Page 161of190 ATTACHMENT 1 temperature shows a modest increase in temperature of around to just below --* Cladding hoop strain goes through a similar transient as shown in Figure 80 but increases from llllltojustoverllllll Time (seconds) Figure 81 Peak Target Pellet Temperature During a 0.0003 Ak/k per Second Reactivity Insertion Page 162of190 ATTACHMENT 1 Time (seconds) Figure 82 Pellet OD and Cladding Temperatures (ID and OD) at Peak Pellet Location During a 0.0003 Ak/k per Second Reactivity Insertion 10.4 Loss of Target Coolant This accident assumes the double-ended break of one of the pipes in the TCS. The design and operation of the TCS are described in Section 3. Pipe breaks can occur in a variety of locations either in or out of the reactor pool. Depending on the location of the break, the loss of coolant will cause either an increase or a decrease in flow at the flow sensors on either pipe leg supplying cooling water to the -T As. The hi/low flow set points are at .. and

of nominal flow and result in a reactor scram and opening of the decay heat removal valves located just above the refueling bridge level. Pump coast-down after pump trip can mitigate the early phase of the transient unless the offset pipe break prevents pump flow from reaching the TA. Pipe breaks in the reactor pool differ from pipe breaks out of the reactor pool (in air) because the decay heat removal valves are assumed to not operate. A pipe break in the pool establishes a natural circulation loop between the TA and reactor pool regardless of whether the decay heat removal valves open. If the break is in the air, a natural circulation loop is only established if the decay heat removal valves open. Otherwise, the break prevents pool water from reaching the TA inlet and establishing a natural circulation loop. Page 163of190 ATTACHMENT 1 The pipe break before the wye affects

T As but a pipe break downstream of the wye at the joint with the flexible pipe has a lower cold leg volume and greater impact on the TA. The relative locations of these pipe breaks are shown in Figure 83. The water in the cold leg of the supply line is almost 18 °F ( 10 0C) cooler than the pool water. The draining of this water through the T As mitigates the transient during the early phase of the loss of coolant accident (LOCA). Figure 83 Pipe Break Locations Out of the Reactor Pool Pipe breaks out of and in the reactor pool are covered in the accident analyses that follow. 10.5 Pipe Break Locations Out of the Reactor Pool The pipe break in air just after the flexible pipe is a more credible pipe break than other locations closer to the TA because the joint is more vulnerable and the remaining pipe is of welded construction. The upper bridge structure protects the pipes where they bend downward into the reactor pool. The break is analyzed to occur in the supply line to TA I because it has the target rod with the highest power density, and the target rod with the highest power in addition to being the TA with the most power. The LOCA for the SGE TEF was modeled using RELAP5 mod3.3 patch 03. RELAP5 was developed by the NRC to analyze thermal-hydraulic transients in pressurized water reactors. It can be used to analyze a variety of geometries and was used to analyze the LOCA and Loss of Flow Accident (LOF A) in Sections 13.2.3 and 13.2.4 of the MURR SAR for relicensing. The RELAP5 model for this accident Page 164of190 ATTACHMENT 1 includes both T As, TCS pump and heat exchanger, and decay heat removal valves. The double-ended guillotine rupture is modeled with three (3) valves -two of the valves connect to the reactor pool on either side of the break, and the third valve connects the two ends of the pipe across the break. Prior to accident initiation, the valves to the reactor pool are closed and the valve across the break is open. When the break is initiated, the valve connecting the two (2) pipe ends is closed and the two (2) valves from each end connected to the reactor pool are opened. All three (3) valves are assumed to completely change position in 0.5 seconds which is a reasonable assumption for a mechanically induced failure that displaces the two (2) ends of the pipe since the piping is not at high pressure. The flow transient caused by a LOCA resulting from a double-ended pipe break is shown in Figure 84. The transient analysis starts with the reactor and T As at 100% power and the TCS at 100% nominal flow to accurately simulate the expected response of the protection system. The decay heat removal valves are assumed to not operate. The break causes an increase in the flow measurement in the pipe supplying TA 2 and a decrease in the flow rate supplying TA 1. The reactor is scrammed at 0.05 seconds when the mass flow rate to TA I is greater than ... A low flow signal occurs shortly thereafter at 0.08 seconds for TA I when its flow rate is less than *. The operator is assumed to immediately trip the TCS water pump on the high flow alarm. CB movement is delayed by 0.15 seconds after the reactor scram signal is generated. The mass flow entering TA I falls quickly because of the pipe break. The flow through TA I remains above for almost 20 seconds. The flow through TA I drops below -Page 165of190 ATTACHMENT 1 Time (sec) Figure 84 Mass Flow Transient During a LOCA in Air Without Decay Heat Removal Valves Opening In TA has the maximum peak power density and -has the maximum rod power. In TA -has the maximum peak power density and .. has the maximum rod power though both rods are lower in power than the rods in *. The heat generation in the target rods is shown in Figure 85. The heat generation drops at around 0.20 seconds when the CBs start inserting. The heat removal from the target rods presented in Figure 87and Figure 89 shows that stored energy removal is significant over the first 10 seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal. Page 166of190 g -e I.! ... .. cu II. ATTACHMENT 1 Time (sec) Figure 85 Target Power During a LOCA in Air Without Decay Heat Removal Valves Opening Coolant temperatures at the inlet and outlet of the -T As are presented in Figure 86. Mass flow out of the T As due to boiling are represented by outlet temperatures between Mass flow back into the T As are represented by the lower bounds of the temperature oscillations around . These flow oscillations also move water and thennal energy to the TA entrance which slowly raises that temperature. Pellet centerline temperatures do not increase during this LOCA because of the reactor scram signal at 0.05 seconds, the. heat capacity, and the water flow through the TA caused by draining of the cooling line. The maximum temperatures of the cladding inner diameter (ID) are shown in Figure 87. The cladding ID temperature increases by prior to CB insertion and afterwards, is lower than nonnal operating temperature. The increases in cladding ID temperature at around -seconds coincide with the decreases in flow rate at those times. Boiling within the TAs keeps the maximum cladding temperature just above saturation temperature. Page 167of190 ATTACHMENT 1 Time (sec) Figure 86 Coolant Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening --.. JI I I ! i 111 ... .... I ; ,,, I I r* I ii I I 11, t \ I 1\ ,.. .... , I I I 11 I ! *--I I I 'I I I I I I I I I ...... J I Time (sec) Figure 87 Maximum Cladding ID Temperatures During a LOCA in Air Without Decay Heat Removal Valves Opening Page 168of190 ATTACHMENT 1 Because the pellet and cladding temperatures are near nomrnl values, all cladding stresses and strains are also nonnal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOCA. If the one of the two decay heat removal valves for each TA were to operate, the consequences of the LOCA in air are even less significant. The flow transient due to a LOCA in air with the decay heat removal valves working is shown in Figure 88. The TA inlet flow is the same as the TA outlet flow and a small natural circulation flow is established after

seconds. TA I flow rate is higher around

seconds due to a small contribution of pump coast-down through the intact supply line to TA I. Time (sec) Figure 88 Mass Flow Transient During a LOCA in Air With Decay Heat Removal Valves Opening Pellet centerline temperatures also do not increase during this LOCA. The maximum temperatures of the cladding ID are shown in Figure 89. The cladding ID temperature increases by prior to control blade insertion and afterwards, is lower than nonnal operating temperature. Natural circulation is sufficient to prevent chugging and cladding temperatures steadily decrease after

seconds. Peak vapor fraction is less than *. Page 169of190 ATTACHMENT 1 Time (sec) Figure 89 Maximum Cladding ID Temperatures During a LOCA in Air With Decay Heat Removal Valves Opening Because the pellet and cladding temperatures are near normal values, all cladding stresses and strains are also normal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOCA. The decay heat removal valves can prevent boiling and chugging in the TA but are not necessary to assure cladding integrity. 10.6 Pipe Break Locations in the Reactor Pool The target cooling lines supplying water to the T As are protected by the upper bridge structure above the pool and the refueling bridge located below the nonnal pool water level. The cooling lines also have lateral supports and are joined to the TA by a flexible joint. It is nearly incredible to postulate a mechanistic failure of these cooling pipes which results in a complete offset rupture within the assumed rupture time of 0.5 sec. If the break occurs near the decay heat removal valves then it is essentially the same as opening the decay heat removal valves. The worst possible break location is the connection between the inlet pipe welded to the target housing and the flexible pipe as shown in Figure 90. Direct mechanical interaction in this area is very unlikely due to the congestion around this elevation. Page 170 of 190 ATTACHMENT 1 Figure 90 Pipe Break Location in the Reactor Pool The LOCA in water used the same RELAP5 model as the LOCAs in air except that the break location was relocated as shown in Figure 90. The break time to complete offset rupture was also kept at

seconds although it is more likely that the break time would be either longer underwater or less than a complete offset rupture. The flow transient caused by a LOCA in water resulting from a double-ended pipe break is shown in Figure 91. The transient analysis starts with the reactor and TA at 100% power and the TCS at 100% nominal flow to accurately simulate the expected response of the protection system. The decay heat removal valves are assumed to not operate though their operation would be ineffective in this accident. The break causes an increase in the flow measurement in the pipe supplying TA I and a decrease in the flow rate supplying TA I. Page 171 of 190 ATTACHMENT 1 Time (sec) Figure 91 Mass Flow Transient During a LOCA in the Reactor Pool The reactor is scrammed at

seconds when the mass flow rate to TA I is greater than ... A low flow signal occurs shortly thereafter at

seconds for TA I when its flow rate is Jess than *. CB movement is delayed by. seconds after the reactor scram signal is generated. The mass flow entering TA I falls quickly because of the pipe break. The flow through TA I drops to zero at around

seconds and experiences several boiling and chugging oscillation for the next few seconds. The flow through TA I remains above zero for about I seconds before it also demonstrates some boiling and chugging oscillations. Natural circulation through TA I after I seconds is sufficient to prevent further boiling and chugging. The heat generation in the target rods is shown in Figure 92. The heat generation drops at around

seconds when the CBs start inserting. Brief periods of transition and film boiling occur at the high heat flux locations of the target rods. The heat removal from the target rods presented in Figure 92 shows that stored energy removal is significant over the first

seconds. Mass flow oscillations due to chugging and boiling are also reflected in the oscillations in heat removal and occur while significant stored energy remains to be removed. Page 172of190 s t: I! ... .. II 0. ATTACHMENT 1 Time (sec) Figure 92 Target Power During a LOCA in the Reactor Pool The maximum temperatures of the cladding ID and coolant entering and exiting TA I are shown in Figure 93. Cladding temperatures start experiencing rapid increases at around. seconds which corresponds to the rapid drop in flow rate. Peak cladding temperatures reach in target rod

and * -in target rod *. Peak cladding temperatures in rods -steadily decline after the reactor scram except for a brief temperature excursion of around -between I and I seconds. A FRAPTRAN analysis of cladding integrity during this LOCA in water transient was performed using minimum and maximum pellet-clad gap tolerances of Cladding OD temperatures as a function of time and axial location were obtained from RELAP5 and used as the boundary condition for the FRAPTRAN analysis. FRAPCON results used in the insertion of excess reactivity transient analysis were used to define the bumup and fission gas release inputs for FRAPTRAN at BOL and EOL conditions. The maximum hoop strain occurs at EOL conditions in target rod

with the maximum gap tolerance of . The cladding temperatures used in the FRAPTRAN analysis for rod

are presented in Figure 94. The maximum hoop strain of-at the peak strain location along with the radial and axial strains at that location are presented in Figure 95. These strains are much less than the 1 % strain criteria so that cladding integrity is maintained. Page 173of190 ATTACHMENT 1 Tlme (sec) Figure 93 Maximum Cladding ID and Coolant Temperatures During a LOCA in the Reactor Pool Figure 94 Cladding OD Temperature Profile in Target Rod. During a LOCA in the Reactor Pool Page 174of190 ATTACHMENT 1 !"* . .,,,., ' .... I ...... .,, '* ,.. ' ,. \ , . ; ....... ,. ' , . \ \ i ; Time (seconds) Figure 95 \ \ \ \ \ \ \ \ Cladding Fractional Strains in Rod

during a LOCA in the Reactor Pool 10. 7 Loss of Target Flow The loss of target flow accident can be initiated by inadvertent valve closures, pump failures, or loss of electrical power. Pipe breaks discussed in the previous section also result in a loss of target flow. Inadvertent valve closure is prevented by securing the valves in the open position using a locking pin since there is no safety requirement associated with the operation of the manual isolation valves and they are only for maintenance purposes. Therefore, the most limiting initiating event is a loss of pump flow (LOPF) which can be caused by either loss of electrical power or pump failure. The LOPF event impacts both T As and includes pump coast-down and fluid momentum to ease the transition from forced flow to natural circulation flow. The TCS has redundant 100% capacity pumps. Only one of the pumps is required to operate and the other pump is an installed spare. If the operating pump fails, there is no automatic switchover to the backup pump. Redundant flow signals in the TCS will initiate protective actions including reactor scram when flow either reduces to

of nominal or increases to .. of nominal. Additional actions taken due to the flow reduction are opening of the decay heat removal valves. The LOPF was modeled using RELAP5 mod3.3 patch 03. The RELAP5 model is the same model used for the LOCA analyses described in Loss of Target Cooling. Page 175of190 ATTACHMENT 1 The flow transient caused by this LOPF is shown in Figure 96. The transient analysis starts with the TCS at I 00% power and 100% nominal flow to accurately simulate the expected response of the protection system. The LOPF accident is assumed to include a loss of secondary flow as one would expect during a loss of site power. The analysis assumes that the decay heat removal valves do not open since they are not an Engineered Safety Feature (ESF). The LOPF causes a decrease in the flow measurement in the pipes supplying . Pump coast-down takes about. seconds to complete. The reactor is scrammed at

seconds when the mass flow rate drops below *. CB movement is delayed by

seconds after the reactor scram signal is generated. As natural circulation progresses, the colder water in the TCS is slowly replaced with water at reactor pool temperature which eventually results in a decline in natural circulation flow at around. seconds. Time (sec) Figure 96 Mass Flow Transient During Loss of Pump Flow Maximum cladding ID temperatures and coolant temperatures are shown in Figure 97. Peak cladding temperatures rise slightly by but then steadily decline after the reactor scram. Coolant exit temperature from TA I rises slightly at around I seconds and again after. seconds. At. seconds the inlet temperature starts to rise as reactor pool water has mixed with the cooler target cooling water. Eventually the TA inlet temperature will reach the maximum pool temperature of 50 °C (120 °F) at which point further increase in temperatures will cease. Because the pellet and cladding temperatures are near nonnal values, all cladding stresses and strains are also nonnal. Therefore, there is no cladding breach and no release of radioactivity associated with this LOPF. The decay heat removal valves can improve the natural circulation cooling in the TA but are not necessary to assure cladding integrity. Page 176of190 ATTACHMENT 1 Time (sec) Figure 97 Maximum Cladding ID and Coolant Temperatures During Loss of Pump Flow 10.8 Mishandling of Target Cartridge or Target Rods This event examines the potential to damage the target cartridge or target rods while being removed from its reflector position to its reactor pool transfer location. Multiple barriers separate the fission products in the target material from potential release locations within the TA. The robust design of the target cartridge and target rods, and careful design of the tooling for handling the cartridge and rods will help prevent a handling accident from occurring while a target cartridge is being moved from its reflector position to its storage and transfer location within the reactor pool, or removing a target rod from the cartridge. This movement occurs no more than twice per week. The movement is carefully planned and does not start until after the reactor is shut down and the target rods are sufficiently cooled so the TCS can be shutdown. Based on additional decay prior to handling, the consequences of mishandling the target cartridge or target rods are bounded by the TEF MHA. 10.9 Loss of Primary Coolant Flow This reactor event can be caused by a variety of initiating events which result is flow stagnation and reversal in the MURR core. For all initiating events other than loss of electrical power, the target cooling water can continue to circulate and cool the TA during this event resulting in no consequences to it. Page 177 of 190 ATTACHMENT 1 10.10 Loss of Primary Coolant This reactor event assumes the double-ended rupture of the largest primary coolant pipe. Low pressure initiates a reactor scram along with primary coolant pump trips, isolation valve closures, and anti-siphon valves opening. Automatic actions ensure that the core remains covered and decay heat is transferred to the pool water. Target cooling water can continue to circulate and cool the TA during this event resulting in no consequences to it. 10.11 Loss of Pool Coolant The loss of pool coolant accident postulates a break which lowers the water level in the reactor pool until the reactor scram set point is reached. Low pool level alanns would activate before reaching that level. The TCS will continue to operate and remove stored energy from the target rods after the reactor scram. The suction line for the TCS is below the reactor scram pool level. When the water has dropped below the suction line, target cooling will no longer be available but the cooling water within the cooling system will continue to drain through the T As. The accident from this point behaves similar to the LOCA in air presented earlier but with much of the stored energy already removed so that no boiling or chugging would occur. Natural circulation and heat transfer through the target cartridge walls would be sufficient to maintain target rod cladding integrity and prevent any release of radioactivity. 10.12 Loss of Offsite Electrical Power This event would cause the reactor to scram, pumps to shut down, and valves to fail to their safe shutdown positions due to their fail-safe design. Loss of electrical power would cause the decay heat removal valves to open thereby replacing the loss of forced flow from the shutdown of the target cooling water pumps with natural circulation flow. Pump coast-down and fluid momentum would provide additional flow during the transition from pump flow to natural circulation flow. This event is the same as the loss of target flow event analyzed earlier. 11. Technical Specifications The new Technical Specifications associated with the SGE TEF are included as Attachment 2 to the License Amendment. 12. Proposed Confirmatory Testing A variety of tests are planned to validate the design and demonstrate operation of the SGE TEF before installation and operation of the experiment in the reactor graphite reflector region. These tests are both at the component and system level, and are designed to validate not only performance and safety at the system level, but to validate analytical modeling of component behavior. Page 178of190 ATTACHMENT 1 12.1 Summary Description of Planned Tests 12.1.1 Target Pellet/Cladding Behavior Irradiation Testing An irradiation test campaign will be performed to demonstrate the safety of the target rods. The chosen location for the test is the National Research Universal (NRU) reactor at Canadian Nuclear Laboratories (CNL), which can provide sufficient neutron flux and similar operating conditions to the nominal target rod design. The test will utilize capsules that replicate the geometry of the target rods. Following the irradiation test, a series of post-irradiation examinations (PIE) will be carried out to verify the performance of the target rod pellets and cladding. CNL will fabricate -LEU irradiation test capsules using components fabricated by GA. Each capsule will contain up to

pellets . The capsule structure will be representative of the full TA's cladding design. This will include a -cladding tube and a nominal gap ofl microns between the cladding and pellets (cold condition). Two ceramic spacers3 will hold the

pellets in the center of the tube to prevent the metal end caps from reaching excessive temperatures. A low-carbon steel retaining spring is included for handling loads. The capsule will be filled with -to improve the gap thermal conductivity between the cladding and pellets. The critical dimensions and attributes are provided in Table 53 and an illustration of the conceptual capsule design is shown in Figure 98, all representative of production target rods. Table 53 Capsule Dimensions and Attributes Cladding ID, mm Cladding wall thickness, mm Pellet density, % TD Pellet OD, mm Pellet height, mm

mass per capsule,* g Pellet enrichment, t% -fill pressure, atm Tube material Ceramic spacer material *Assumes 20 pellets per capsule at 95% theoretical density. tValue determined by CNL neutronics analysis. ----*

I -3 The peak neutron flux in the NRU reactor is essentially uniform over the length of the capsule. A ceramic spacer is therefore included in the test capsule to prevent end effect over heating of the capsule. Page 179of190 ATTACHMENT 1 Figure 98 Conceptual Design of Test Capsule The capsules are designed to interface with CNL' s existing equipment for handling nuclear targets used for the production of medical isotopes in the NRU reactor. The flow channel for the target capsules is formed long rod-like structure which replaces a fuel element. This structure contains four internal channels, each of which house an assembly of -target elements linked end over end in a straight line, referred to as a stringer. The ends of the test capsules incorporate features which allow them to be incorporated into the stringer assemblies. For the irradiation test, each stringer will consist of -. By utilizing the existing stringer design, the test capsules will allow for flexible handling, as each stringer can be removed separately. The duration of the test irradiation period in the reactor is , the maximum allowable irradiation time for target rods in MURR. The coolant flow conditions in the NRU reactor are similar to those in MURR. The coolant temperature ranges from . Since there are multiple available irradiation locations, the target position was chosen to ensure that the targets reach the maximum design pellet power density of-* Given the highly thermalized neutron spectrum of the NRU reactor, the pellet enrichment was reduced to reach the desired power density. By matching the power density, coolant temperature and flow velocity, the test ensures that the irradiation cond.itions are sufficiently similar to the design condition in MURR. Page 180of190 ATTACHMENT 1 Since FRAPCON/FRAPTRAN analysis shows that relocation strain occurring at shutdown contributes to the closure of the pellet-cladding gap, the test will incorporate multiple shutdown/restart cycles. -will be simulated by removal of the test capsules from the flux field (the targets are compatible with online handling). The test schedule is given below in Table 54. Test -Capsules to be Subjected to Irradiation Tests for Pellet/Cladding Behavior Table 54 Irradiation Test Schedule Tasks Irradiations Post-irradiation Examinations (PIE) Test Report Preparation Start ---Following the irradiation, the capsules will be subjected to numerous PIE tests. End ---irradiation, the pellets may deform due to thermal strains, including "hour-glassing" and cracking, and release gaseous fission products. A visualization of potential pellet deformation during irradiation is shown in Figure 99. This deformation and gas build-up can lead to both mechanical and chemical interactions between the cladding and pellets, and verifying that these phenomena do not compromise the integrity of the cladding is important to the safety and operation of the target rods. Figure 99 Visualization of Potential Pellet Deformation over Course of Irradiation (not to scale) Page 181 of 190 ATTACHMENT l The following PIE tests will be perfonned on the capsules following the irradiation period: ---Data collected from the PIE will be provided to the NRC as soon as it becomes available. While the test is intended to verify GA's predictions of target rod behavior, the test will be considered successful as long as the test capsule remains hermetically sealed prior to destructive examination. 12.1.2 Critical Heat Flux Tests The SGE target rods operate at very high power densities to maximize isotope yield. Analysis of the heat transfer in the target systems shows that cladding wall temperatures in the highest heat flux regions of the target exceed the coolant saturation temperature, leading to some subcooled nucleate boiling. The analysis predicts that the vapor content due to this boiling is confined to the surface of the target rods and is negligible in volume, and the NRC-accepted correlations show that the CHFR for the system will continue to provide sufficient margin. While the CHF with the various correlations shows that the CHFR for the target rod cartridge provides sufficient margin, the safest approach is to verify that there are Page 182of190 ATTACHMENT 1 adequate margins in the CHFR via an experiment that shows significant margin exists between the system's maximum operating heat flux and the CHF for the configuration. The overall objectives of the test are as follows:

Demonstrate that for the system's maximum design heat flux, the cooling flow remains in the subcooled nucleate boiling regime, with minimal vapor generation at the wall.

Experimentally determine the CHF for the system geometry and flow conditions, showing that sufficient safety margin is maintained. This test is being conducted at the University of Wisconsin. section design is shown in Figure 100 and the conceptual section of the test rig minus the instrumentation is shown in Figure 101. Heating is provided by direct resistive heating of an -tube. The power supply is rated to 100 kW and is capable of supplying a heat flux at the wall of -* Instrumentation to measure pressure drop, coolant temperature, Inconel tube temperature, and power supplied are all incorporated into the test section design. Coolant temperature and pressure in the test rig are adjustable to match target rod design conditions. Since the resistive heating provides a flat heat profile, the length of the heat tube was reduced by the peaking factor so that the total power was preserved while operating at the maximum heat flux. The first test objective will be achieved by operating the flow loop while providing sufficient power to the heating rod to reach the desired heat flux. Nominal and worst-case conditions will be used. Cladding surface temperature will be calculated from internal thermocouple measurements, power supply output, and heating tube geometry. The bubble generation at heater surface will be captured with high speed camera. The second test objective will be achieved by slowly ramping the power to the heater until DNB occurs. DNB is detectable as an excursion event on the heater internal thermocouple. Coolant temperature will be adjusted to ensure that local conditions at the end of the heater where DNB occurs match the design temperature of the coolant at the maximum heat flux location in the target cartridge. The test schedule is given below in Table 55. Page 183of190 Test Critical Heat Flux Tests ATTACHMENT 1 Table 55 Critical Heat Flux Test Schedule Tasks Complete Assembly and Instrumentation of Test Rig and Perfonn Initial Shakedown Testing Perfonn Flow Characterization and CHFR Tests Complete Final Report Figure 100 Start ---Conceptual Schematic of CHF Test Flow Section Page 184of190 End ---

ATTACHMENT 1 Figure 101 Uninstrumented Critical Heat Flux Test Section in Low Pressure Flow Loop Rig 12.1.3 System Integration and Cooling Flow Tests The purpose of the system integration is to verify the installation and removal procedures for the target system components (target housing, target cartridge, target rods, and target loading and unloading station) and TCS piping upstream of the target housing. Operation of the pick-up tooling for the target rods and cartridges will also be demonstrated. In addition, after system integration is complete, flow and pressure drop tests will be performed to verify the functionality of the WCM and to confinn the pressure drop through the TA that was obtained by analysis. A representation of these systems will be assembled and tested at GA facilities. Figure 102 illustrates the test components, which include a surrogate pool, target system, and TCS. Page 185of190 ATTACHMENT 1 Figure 102 GA Test Setup Conceptual Arrangement The surrogate pool will be a fiberglass tank; . This height allows realistic underwater simulation of the system integration procedures. The TCS piping in the surrogate pool will be representative of the piping and pipe support structure that will be used at MURR. The goal is to mimic the attachment points and structural interferences as closely as possible. The full scale production system at MURR consists of two target systems, however only one target system will be tested in this configuration at GA's test facilities. The test specifications and a comparison of the out-of-reactor test pool and the MURR pool are shown in Table 56. Page 186of190 ATTACHMENT 1 Parameters Pool Diameter Pool Depth Distance from Bottom of Target to Nominal 0 eratin Water Level WCM Inlet Pipe Inner Diameter Number of Target Assemblies WCM Outlet Pipe Inner Diameter Water Chemistry The test schedule is shown in Table 57. Table 56 Testing Specifications GA Test Pool **

Specifications * .. ---I -Table 57 System Integration Test Schedule Tes.t Tasks. TA Installation Test MURR fool Specifications ----I -Start . En:d --TA Pressure Drop Test --System Integration and 1----------------+------+-------1 Cooling Flow Tests Target Rod & Cartridge --Installation/Removal Test Test Report Preparation --Page 187of190 ATTACHMENT 1 13. References 1. "Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes", Tenth International Symposium, ASTM STP 1245, 1994 pp. 549-558. 2. Nekhamkin, Y., Hasan, D., Elias, E. "Zirconium Ignition in an Exposed Fuel Channel," Conference on Reactor Physics and Technology II, Israel, pp. 45-48, February 2014. 3. [ASTM STP 1245] "Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes", Tenth International Symposium, ASTM STP 1245, 1994, pp 549-558. 4. 2015 "Primary Radiation Damage Cross Sections, https://www-nds.iaea.org/CRPdpa/, 2015. 5. [304441M00014/B] "Maximum Neutron Damage of Al6061, Zircaloy-4 and SS316L" GA Document No. 30441M00014/B. General Atomics Proprietary Infonnation. 6. [King, 1973] King, T. T., A. Jostons and K. Farrell, "Neutron irradiation damage in a precipitation-hardened aluminum alloy," Effects of Radiation on Substructure and Mechanical Properties of Metals and Alloys, ASTM STP 529, American Society for Testing and Materials, 1973, pp. 165-180. 7. [Jin 2015] Jin, H.J. and T.K. Kim, "Neutron Irradiation Perfonnance of Zircaloy-4 under Research Reactor Operating Conditions," Annals of Nuclear Energy, 25 (2015) pp. 309-315. 8. [IAEA-TECDOC-1496] Thermal Physical Properties Database of Materials for Light Water and Heavy Water Reactors, IAEA-TECDOC-1496, June 2006. 9. [Byun 1996] Byun, T. S. and N. Hashimoto, "Strain hardening and long-range internal stress in the localized deformation of irradiated polycrystalline metals," Journal of Nuclear Materials 354 (2006) 123-1304-28. 10. Pawel, J. E. et. al., Irradiation perfornmnce of stainless steels for ITER applications," Journal of Nuclear Materials 239 (1996) 126131. 11. University of Missouri Research Reactor Safety Analysis Report, submitted to the U.S. Nuclear Regulatory Commission August 2006. 12. [LANL 2003] "MCNP -A General Monte Carlo N-Particle Transport Core, Version 5," LA-UR-03-1987, Los Alamos National Laboratory, April 2003. 13. [Conlin 2013] Conlin, J. L. et al., "Continuous Energy Neutron Cross Section Data Tables Based upon ENDF/B-VIl.l," LA-UR-13-20137, Los Alamos National Laboratory, Feb. 2013. 14. [Parsons 2012] Parsons, D. K., Conlin, J. L, "Release of Continuous Representation for S(a,p) ACE Data," LA-UR-12-00800, Los Alamos National Laboratory, February 2012. 15. [Kutikkad 2015] Kutikkad, K., Private Communication, University of Missouri Research Reactor, July 2015. Page 188of190 ATTACHMENT 1 16. [Peters 2016] Peters N., Private Communication, University of Missouri Research Reactor, May 2016. 17. [Kiedrowski 201 OJ Kiedrowski, B. C. et al., "MCNP5-l.60 Feature Enhancement & Manual Clarifications," LA-UR-10-06217, Los Alamos national Laboratory, 2010. 18. M. C. White, "Photoatomic Data Library MCPLIB04: A New Photoatomic Library Based On Data from ENDF/B-VI Release 8," LA-UR-03-1019, Los Alamos National Laboratory, February 2003. 19. M. C. White, "Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code," LA-13744-T, Los Alamos National Laboratory, July 2000. 20. J. F. Briesmeister, Ed., "MCNPTM -A General Monte Carlo N-Particle Transport Code," LA-13709-M, Los Alamos National Laboratory, March 2000. 21. J. W. Durkee, Jr. et al., "The MCNP6 Delayed-Particle Feature," LA-UR-12-00283, Los Alamos National Laboratory, March 2012. 22. D. B. Pelowitz, Ed., "MCNP6TM User's Manual Version 1.0," LA-CP-13-00634, Los Alamos National Laboratory, May 2013. 23. [30441R00033 2016] GA Document No. 30441R00033/B. "Forced Convection Cooling of High Power Density Nuclear Target Rod with Two-Phase Considerations" (December 2016) General Atomics Proprietary Infonnation. 24. [Del Valle 1985] V.H. Del Valle, D.B.R. Kenning, Subcooled flow boiling at high heat flux, Int. J. Heat Mass Transfer, 28 (1985), pp. 1907-1920). 25. [Jens 1951] Jens, W.H., and Lottes, P.A, "Analysis of heat transfer, burnout, pressure drop and density data for high pressure water", ANL-4627, 1951. 26. [Unal 1976] H.C. Unal, Maximum Bubble Diameter, Maximum Bubble-Growth Time and Bubble-Growth Rate During the Subcooled Nucleate Flow Boiling of Water up to 17. 72 MN/m"2, Int. J. Heat Mass Transfer, 19 (1976), pp. 643-649. 27. [Chen 1966] Chen, J. C. "Correlation for Boiling Heat Transfer to Saturated Liquids in Convective Flow," in Ind. Eng. Chem. Proc. Dev., 5, 322, 1966. 28. [Glasstone & Sesonske, 1967] S. Glasstone and A. Sesonske, Nuclear Reactor Engineering, Chapter 6, "Heat Transfer to Boiling Liquids: Surface and Volume Boiling", Sec. 6.129-6.149, Van Nostrand Reinhold [1967]. 29. [Bernarth 1960] Bernath, Louis, "A Theory of Local-Boiling Burnout and its Application to Existing Data", Chemical Engineering Progress Symposium Series, NO. 30, Vol 56, p.95-116, 1960. 30. [FRAPTRAN 2011] FRAPTRAN 1.5: "A Computer Code for the Transient Analysis of Oxide Fuel Rods", NUREG/CR-7023, PNNL-19400, Vol. 1, Rev. 1May2014. Page 189of190 ATTACHMENT 1 31. [Groeneveld 2007] Groeneveld, D.C., et al, The 2006 CHF Look-Up Table, Nuclear Engineering and Design 23 7 (2007) p. 1909-1922. 32. [Padoussis 2004] ASME Boiler and Pressure Vessel Code,Section III, Div 1, Appendix N, N-1345.1, p. 338, 2004. 33. [Fanner 2006] Farmer, M.T., Hoffman, E.A., Pfeiffer, P.F., Therios, 1.U, Wei, T.Y.C, Generation IV Nuclear Energy System Initiative Pin Core Subassembly Design, ANL-GENIV-070, April , 2006, p. 57-59. 34. [30441R00035 2016] "Cooling of the MURR Beryllium Reflector" GA Document No. 30441R00035/A (December 2016) General Atomics Proprietary Information. 35. GA Doc. No. 30441R00022-Moved to Applicable Documents Table 3-2. 36. "Decay Heat Power in Light Water Reactors," ANSI/ANS-5.1-2014, published by the American Nuclear Society. 37. "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel," ANSI/ANS-5.4-2011, published by the American Nuclear Society. 38. [30441R00022 2016] "Source Term Analysis" GA Document No. 30441R00022/B (December 2016) General Atomics Proprietary Infonnation. Page 190of190 ATTACHMENT 2 Implementation of the Selective Gas Extraction (SGE) Target Experimental Facility (TEP) at the University of Missouri Research Reactor (MURR) will require minor changes to a number of existing MURR Technical Specifications (TS) as well as new TSs. I. The following current TSs will require revision: 1. TS 3 .2.g -Reactor Control and Reactor Safety Systems 2. TS 3.8.f-Experiments 3. TS 3.8.n -Experiments 4. TS 3.8.r-Experiments 5. TS 3.8.t-Experiments II. Proposed changes to these Specifications are as follows (changes are in bold text): Specifications: 3.2.g There will be five (5) new reactor safety system instrument channels (22 through 26) and four (4) new corresponding notes (6 through 9) and revised bases. See attached TS pages. 3.8.f Each fueled experiment shall be limited such that the total inventory of iodine-131 through iodine-135 in the experiment is not greater than 150 Curies and the maximum strontium-90 inventory is no greater than 300 millicuries. Exception: SGE target rods. 3.8.n Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the containment building atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity when averaged over a year will not exceed the limits of 10 CPR 20. Exception: Fueled experiments that produce iodine-131 through iodine-135, SGE target rods that produce iodine-131 through iodine-135, and non-fueled experiments that are intended to produce iodine-131 (See Specifications 3.8.f, 3.8.h and Amendment No. X). 3.8.r Cooling shall be provided to prevent the surface temperature of a submerged irradiated experiment from exceeding the saturation temperature of the cooling medium. Exception: SGE target rods. 3.8.t The maximum temperature of a fueled experiment shall be restricted to at least a factor of two (2) below the melting temperature of any material in the experiment. First-of-a-kind fueled experiments shall be instrumented to measure temperature. Exception: SGE target rods. Bases: f. Specification 3.8.f restricts the generation of hazardous materials to levels that can be handled safely and easily. With the exception of SGE target rods, analysis of fueled experiments containing a greater inventory of fission products has not been completed, 1of3 ATTACHMENT 2 and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR and Section 10.1 of Amendment No. X). r. Specification 3.8.r is intended to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation (Ref. Sections 4.5 and 13.2.6 of the SAR and Sections 5 and 6 of Amendment No. X). t. Specification 3.8.t is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure. Amendment No. X provides the safety analysis shows why the SGE target rod can be excepted from Specification 3.8.t. III. The following five (5) new proposed TS definitions will be added: 1.37 Selective Gas Extraction (SGE) Target Assembly -The Selective Gas Extraction (SGE) Target Assembly consists of (1) a water inlet section, (2) a target housing, (3) a lower plenum, (4) a target cartridge, (5) an outlet diffuser, and (6) a target cartridge locking mechanism. 1.38 Selective Gas Extraction (SGE) Target Cartridge -The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and ( 4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, (4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and (5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation. 1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target cartridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-enriched uranium dioxide pellets nominally enriched to 19.75% in the isotope uranium-235 with an active length of 600+/-3 millimeters. 2 of3 ATTACHMENT 2 IV. The following new TS Sections will be added (see attached pages): 3.11. Selective Gas Extraction Target Experimental Facility (Limiting Conditions of Operations) 4.11 Selective Gas Extraction Target Experimental Facility (Surveillance Requirements) 5.7 Selective Gas Extraction Target Experimental Facility (Design Features) 3 of3 1 DEFINITIONS -Continued UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFIC A TIO NS Docket No. 50-186, License No. R-103 the principal physical baITiers which guard against the uncontrolled release of radioactivity. 1.35 Scram Time -Scram time is the elapsed time between the initiation of a scram signal and insertion of the shim blades to the 20% withdrawn position. 1.36 Secured Experiment -A secured experiment is any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restrai¢ng forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are normal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions. 1.37 Selective Gas Extraction (SGE) Target Assembly -The Selective Gas Extraction (SGE) Target Assembly consists of (1) a water inlet section, (2) a target housing, (3) a lower plenum, (4) a target cartridge, (5) an outlet diffuser, and ( 6) a target cartridge locking mechanism. 1.38 Selective Gas Extraction (SGE) Target Cartridge -The Selective Gas Extraction (SGE) Target Cartridge is designed to (1) position and support the SGE target rods, (2) provide a cooling passage for the SGE target rods, (3) reduce neutron flux peaking at the axial center by the use of a neutron absorber, and ( 4) mix and guide the cooling water outlet flow. 1.39 Selective Gas Extraction (SGE) Target Experimental Facility (TEF) -The Selective Gas Extraction (SGE) Target Experimental Facility (TEF) is designed to (1) provide forced cooling to the target rods during normal operation, (2) provide natural circulation cooling during shutdown periods, (3) transfer and reject heat to the secondary coolant system, ( 4) maintain sufficient cooling flow and temperature conditions to ensure a Critical Heat Flux Ratio of greater than 2.0, and (5) provide instrumentation to assure cooling flow rates, temperatures, and pressures are within specified conditions for operation. 1.40 Selective Gas Extraction (SGE) Target Housing -The Selective Gas Extraction (SGE) Target Housing is designed to (1) direct the flow of cooling water, (2) provide structural support, and (3) position the target caiiridge within the graphite reflector region. 1.41 Selective Gas Extraction (SGE) Target Rod -The Selective Gas Extraction (SGE) Target Rods consist of upper and lower Zircaloy-4 end caps, Zircaloy-4 cladding, a stainless steel spring, and low-emiched uranium dioxide pellets A-X 1 DEFINITIONS -Continued UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 nominally enriched to 19.75% in the isotope uranium-235 with an active length of 600+/-3 millimeters. 1.36 Secured Experiment -A secured experiment is any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are nmmal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions. 1.37 Senior Reactor Operator -A senior reactor operator is an individual who is licensed to direct the activities of reactor operators and manipulate the controls of a reactor. 1.38 Shim Blade (Rod) -A shim blade (rod) is a high worth control blade (rod) used for coarse adjustments in the neutron density and to compensate for routine reactivity losses. The shim blade (rod) is magnetically coupled to its drive mechanism allowing it to scram when the electromagnet is de-energized. The shim blade (rod) also provides rod run-in functions. 1.39 Shall, Should, and May -The word "shall" is used to denote a requirement; the word "should" is used to denote a recommendation; and the word "may" is used to denote permission, neither a requirement nor a recommendation. 1.40 Shutdown Margin -Shutdown margin is the minimum shutdown react1v1ty necessary to provide confidence that the reactor can be made subcritical by means of the control and reactor safety systems starting from any permissible operating condition and with the most reactive shim blade and the regulating blade in the fully withdrawn positions, and that the reactor will remain subcritical without further operator action. 1.41 Surveillance Intervals -Surveillance intervals are the maximum allowable intervals established to provide operational flexibility and not reduce frequency. Established frequencies shall be maintained over the long term. The surveillance interval is the time between a check, test or calibration, whichever is appropriate to the item being subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following: a. Biennial -interval not to exceed 2.5 years. b. Annual -interval not to exceed 15 months. A-X

,.i / 1 DEFINITIONS -Continued UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 c. Semiannual -interval not to exceed 7.5 months. d. Quarterly -interval not to exceed 4 months. e. Monthly -interval not to exceed 6 weeks. f. Weekly-interval not to exceed 10 days. g. Daily -interval not to exceed 1 calendar day. h. Within a shift -interval not to exceed the reactor shift. 1.42 True Value -The true value is the actual value of a parameter. 1.43 Unscheduled Shutdown -An unscheduled shutdown is defined as any unplanned shutdown, that occurs after all "Blade Full-In Lights" have cleared, caused by actuation of the reactor safety system, rod run-in system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations. 1.44 Unsecured Experiment -An unsecured experiment is any experiment which is not secured as defined by Specification 1.36, or the moving parts of secured experiments when they are in motion. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued 18. 19. 20. 21. 22. 23. 24. 25. 26. Reactor Safety System Number Required (N) Instrument Channel Mode I Mode II Mode III Set Point Scram as a result of Facility Evacuation 1 1 1 actuating the facility evacuation system Scram as a result of Reactor Isolation 1 1 1 actuating the reactor isolation system Manual Scram 1 1 1 Push button on Control Console Scram as a result of removing the center Center Test Hole i5) i5) i5) test hole removable experiment test tubes or strainer SGE Target Assembly i6) 0(8) 0(8) 95 gpm (Min) 'A' Coolant Low Flow SGE Target Assembly 2(6) 0(8) 0(8) 129 gpm (Max) 'A' Coolant High Flow SGE Target Assembly i7) 0(8) 0(8) 95 gpm (Min) 'B' Coolant Low Flow SGE Target Assembly 2(7) 0(8) 0(8) 129 gpm (Max) 'B' Coolant High Flow SGE Heat Exchanger Outlet Water 2(9) 0(8) 0(8) 105 °F (Max) Temperature (!) These Instrument Channels are not required when in Mode III operation below 50 kW in natural convection cooling (natural convection flange and pressure vessel cover removed). These Instrument Channels are required when in Mode III operation with forced cooling. C2l Flow orifice (instrumentation displayed in gpm) or heat exchanger (instrumentation displayed in psi) in each operating heat exchanger leg corresponding to the flow value in the table. C3l Core (instrumentation displayed in psi) corresponding to the core flow value in the table. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued C4) Trip pressure is that which corresponds to the pressurizer pressure indicated in the table with normal primary coolant flow. (S) Not required if reactivity worth of the center test hole removable experiment sample canister and its contents or the strainer is less than the reactivity limit of Specification 3.8.b. This safety function shall only be bypassed with specific authorization from the Reactor Manager. (6) Not required if SGE target assembly 'A' is not in operation. This safety function shall only be bypassed with specific authorization from the Reactor Manager. C7) Not required if SGE target assembly 'B' is not in operation. This safety function shall only be bypassed with specific authorization from the Reactor Manager. (S) The SGE Target Experimental Facility shall only be operated during Mode I operation; therefore, not required for Mode II and III operation. This safety function shall only be bypassed with specific authorization from the Reactor Manager. C9) Not required if the SGE Target Experimental Facility is secured. This safety function shall only be bypassed with specific authorization from the Reactor Manager. h. The following reactor control interlocks shall be operable whenever the reactor is in operation. Minimum Numbers Interlock Function Prevents the control rods from Rod Withdrawal being withdrawn unless the 1. Prohibit control system logic functions 1 listed in the Bases have been satisfied Prevents placing the reactor in Automatic Control automatic control unless the 2. Prohibit control system logic functions 1 listed in the Bases have been satisfied Bases: a. Specification 3.2.a ensures that the normal method of reactivity control is used during reactor operation (Ref. Section 4.5 of the SAR). A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued b. Specification 3.2.b provides a restriction on the maximum neutron flux tilting that can occur in the core to ensure the validity of the power peaking factors described in Section 4.5 of the SAR. c. Specification 3 .2.c assures prompt shutdown of the reactor in the event a scram signal is received as analyzed in Section 13.2.2 of the SAR. The 20% level is defined as 20% of the shim blade full travel as measured from the fully inserted position. Below the 20% level, the fall of the shim blade is cushioned by a dashpot assembly. Approximately 91 % of the shim blade total worth is inserted at the 20% level. d. Specification 3 .2.d limits the rate of reactivity addition by the regulating blade to provide for a reasonable response from operator control (Ref. Section 4.5 of the SAR). This Specification is based on a regulating blade total reactivity worth limit of 6.0 x 10-3 (Specification 5.3.d) and a regulating blade travel speed of 40 inches per minute. e. Specification 3.2.e assures that power increases caused by control rod motion will be safely terminated by the reactor safety system. The continuous control rod withdrawal accident is analyzed in Section 13.2.2 of the SAR. Based on a total shim blade reactivity worth of 0.1838 and a maximum shim blade travel speed of 2 inches per minute in the inward direction, the maximum insertion rate of negative reactivity would be 2.4 x 10*4 Based on a maximum shim blade travel speed of 1 inch per minute in the outward direction, the maximum insertion rate of positive reactivity would be 1.2 x 10-4 (or 2.1 x 10-4 at the peak worth region of the shim blade bank). Both values are less than Specification 3.2.e limit of 3.0 x 104 The continuous rod withdrawal accident analyzed in Addenda 1 and 5 to the MURR Hazards Summary Report used reactivity insertion rates of 2.78 x 10*4 and 3.0 x 10-4 respectively. f. The specifications on high power level and short reactor period are provided to introduce shim blade insertion on a reactor transient before the reactor safety system trip is actuated. The low pool level rod run-in provides assurance that the radiation level from direct core radiation above the pool will not exceed 2.5 mR/h (Ref. Section 11.1.5.l of the SAR). The vent tank low level rod run-in prevents reactor operation with a vent tank level which could result in the introduction of air into the primary coolant system (Ref. Section 9.13 of the SAR). A-X .

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued The anti-siphon system high level rod run-in provides assurance that the introduction of air to the invert loop is sufficiently rapid to prevent a siphoning action following a rupture of the primary coolant piping (Ref. Section 6.3 of the SAR). The rod not-in-contact with magnet rod run-in assures the reactor cannot be operated in violation of Specification 3.2.b due to a dropped shim blade. The specification on the truck entry door prohibits reactor operation without the door's contribution to containment integrity as required by Specification 3.4.a. The regulating blade rod run-ins ensure termination of a transient which, m automatic control, is causing a rapid insertion of the regulating blade. g. The specifications on high power level, primary coolant flow, primary coolant pressure and reactor inlet water temperature provide for the limiting safety system settings outlined in Specifications 2.2.a, 2.2.b and 2.2.c. In Mode I and Mode II operation, the core differential temperature is approximately 17 °F; therefore, the reactor outlet water temperature scram set point at 175 °F provides a backup to the high reactor inlet water temperature scram. The core differential pressure scram provides a backup to the primary coolant low flow scrams. The reactor period scram assures protection of the fuel elements from a continuous control blade withdrawal accident as analyzed in Section 13.2.2 of the SAR. With the reflector plenum natural convection valve V547 in the open position and a pool coolant flow rate at 850 gpm, the pool coolant low flow scram assures the adequate cooling of the reactor pool, reflectors, control rods, and the flux trap (Ref. Section 5.3.5 of the SAR). The reflector high and low differential pressure scram provides a backup to the low pool coolant flow scram. The pressurizer high pressure scram provides assurance that the reactor will be shut down during a high pressure transient before the relief valve set point or the pressure limit of the primary coolant system is reached as analyzed in Section 13.2.9.4 of the SAR. The pressurizer low level scram provides assurance that the reactor will be shut down on a loss of coolant accident before pressurizer level decreases sufficiently to introduce nitrogen gas into the primary coolant system. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3 .2 Reactor Control and Reactor Safety Systems -Continued The pool water low level scram assures that the radiation level above the reactor pool from direct core radiation remains below 2.5 mR/h (Ref. Section 11.1.5.1 of the SAR). The reactor scrams caused by the primary and pool coolant isolation valves (V507A/B and V509) leaving their full open position provide the first line of protection for a loss of flow accident (in their respective system) initiated by an inadvertent closure of the isolation valve(s). The power level interlock (PLI) scram provides assurance that the reactor cannot be operated with a power level greater than that authorized for the mode of operation selected on the Power Level Switch. The PLI scram also provides the interlocks to assure that the reactor cannot be operated in Mode I with a primary or pool coolant low flow scram bypassed. The facility evacuation and reactor isolation scrams provide assurance that the reactor is shut down for any condition which initiates or leads to the initiation of a facility evacuation or an isolation of the reactor containment building. The manual scram provides assurance that the reactor can be shut down by the operator if an automatic function fails to initiate a reactor scram or if the operator detects an impending unsafe condition prior to the initiation of an automatic scram. The center test hole scram provides assurance that the reactor cannot be operated unless the removable experiment sample canister or the strainer is inserted and latched in the center test hole. This is required anytime the reactivity worth of the center test hole removable experiment sample canister and the contained experiments or the strainer exceeds the limit of Specification 3.8.b (Ref. Section 13 .2.2 of the SAR). The center test hole scram may be bypassed if the total reactivity worth of the removable experiment sample canister and the contained experiments or the strainer does not exceed the limit of Specification 3.8.b and is authorized by the Reactor Manager. The SGE target assembly coolant high and low flow scrams provides assurance that the reactor will be shut down on a loss of target coolant or forced cooling flow (Ref. Section 10 of Amendment No. X). The SGE heat exchanger high outlet water temperature scram provides assurance that the reactor will be shut down on a loss of secondary cooling flow to the SGE Target Experimental Facility. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.2 Reactor Control and Reactor Safety Systems -Continued h. Specification 3.2.h assures that certain system conditions have been met prior to conducting a reactor startup (Master Control Switch 1S1 in the "ON" position; No Nuclear Instrument anomaly; Shim rods bottomed and in contact with their electromagnets; Source range level indication greater than 20 cps or intermediate range level recorder indication greater than 2 x 1 o-5% power; and Thermal Column door closed) or placing the reactor in automatic control at power (Reactor period as indicated by Inte1mediate Range Channels 2 and 3 greater than 3 5 seconds; Indicated reactor power level greater than the "auto control prohibit" set point on the wide range neutron flux monitor recorder; Regulating blade position greater than 60% withdrawn; and Range Selector Switch 1 S2 in the 5-kW red scale position or above) (Ref. Sections 7.5.3.1 and 7.5.4 of the SAR). A-X 3.8 Experiments Applicability: UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 This specification applies to all experiments which directly utilize neutrons or other radiation produced by the reactor. Radioactive sources shall meet the requirements for experiments. Objective: The objective of this specification is to prevent an accident which would jeopardize the safe operation of the reactor or would constitute a hazard to the safety of the facility staff and general public. Reactivity Limits Specification: a. The absolute value of the reactivity worth of each secured removable experiment shall be limited to 0.006 b. The absolute value of the reactivity worth of all experiments in the center test hole shall be limited to 0.006 c. Each movable experiment or the movable parts of any individual experiment shall have a maximum absolute reactivity worth of 0.001 d. The absolute value of the reactivity worth of each unsecured experiment shall be limited to 0.0025 e. The absolute value of the reactivity worth of all unsecured experiments which are in the reactor shall be limited to 0.006 Materials Specification: f. Each fueled experiment shall be limited such that the total inventory of iodine-131 through iodine-135 in the experiment is not greater than 150 Curies and the maximum strontium-90 inventory is no greater than 300 millicuries. Exception: SGE target rods. g. Fueled experiments containing inventories of iodine-131 through iodine-13 5 greater than 1.5 Curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy the requirements of Specification 3.8.s or be vented to the facility ventilation exhaust stack through high efficiency particulate air (HEP A) and charcoal filters which are continuously monitored for an increase in radiation levels. h. Each non-fueled experiment that is intended to produce iodine-131 shall be limited such that the inventory of iodine-131 is not greater than 150 Curies. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments -Continued 1. Explosive materials shall not be iITadiated nor shall they be allowed to generate in any experiment in quantities over 25 milligrams of TNT-equivalent explosives. Explosive materials shall be limited to a total quantity of 100 milligrams of equivalent explosives in the reactor containment building. J. CoITosive materials shall be doubly encapsulated in conosion-resistant containers to prevent interaction with reactor components or pool water. Should a failure of the encapsulation occur that could damage the reactor, then the potentially damaged components shall be removed and inspected. k. Cryogenic liquids shall not be used in any experiment within the reactor pool. 1. Fluids shall only be utilized in beamport loop experiments and shall be of types which will not chemically react in the event of leakage and shall be maintained at pressure and temperature conditions such that the integrity of the beam tube will not be impaired in the event of loop rupture. m. The normal operating procedures shall include controls on the use or. exclusion of coITosive, flammable, and toxic materials in experiments or in the reactor containment building. These procedural controls shall include a current list of those materials which shall not be used and the specific controls and procedures applicable to the use of conosive, flammable, or toxic materials which are authorized. Failure and Malfunctions Specification: n. Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the containment building atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity when averaged over a year will not exceed the limits of 10 CFR 20. Exception: Fueled experiments that produce iodine-131 through iodine-135, SGE target rods that produce iodine-131 through iodine-135, and non-fueled experiments that are intended to produce iodine-131 (See Specifications 3.8.f, 3.8.h and Amendment No. X). o. Experiments shall be designed and operated so that identifiable accidents such as a loss of primary coolant flow, loss of experiment cooling, etc., will not result in a release of fission products or radioactive materials from the experiment. p. Experiments shall be designed such that a failure of an experiment will not lead to a direct failure of another experiment, a failure of a reactor fuel element, or to interfere with the action of the reactor safety and reactor control systems or other operating components. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments -Continued q. No experiments shall be placed in the reactor pressure vessel or water annulus surrounding the center test hole other than for reactor calibration. r. Cooling shall be provided to prevent the surface temperature of a submerged irradiated experiment from exceeding the saturation temperature of the cooling medium. Exception: SGE target rods. s. Irradiation containers to be used in the reactor, in which a static pressure will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2). t. The maximum temperature of a fueled experiment shall be restricted to at least a factor of two (2) below the melting temperature of any material in the experiment. First-of-a-kind fueled experiments shall be instrumented to measure temperature. Exception: SGE target rods. Other Specification: u. Only movable experiments in the center test hole shall be removed or installed with the reactor operating. All other experiments in the center test hole shall be removed or installed only with the reactor shutdown. Secured experiments shall be rigidly held in place during reactor operation. v. Non-fueled experiments that are intended to produce iodine-131 shall be processed in hot cells that are vented to the exhaust stack system through charcoal filters which are continuously monitored for an increase in radiation levels. Bases: a. Specification 3.8.a provides assurance that any inadvertent insertion/removal or credible malfunction of a secured removable experiment would not introduce positive reactivity whose consequences would lead to radiation exposures in excess of the 10 CFR 20 limits. The step reactivity insertion is analyzed in Section 13.2.2 of the SAR. b. The reactivity worth of experiments in the center test hole is limited by Specification 3.8.b such that the introduction of the maximum reactivity worth of all experiments would not result in damage to the fuel plates as analyzed in Section 13.2.2 of the SAR. c. Specification 3.8.c provides assurance that the movement of movable experiments or movable parts of any experiment will not introduce reactivity transients more severe than one that can be controlled without initiating a reactor safety system action as analyzed in Section 13.2.2 of the SAR. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments -Continued d. Specification 3.8.d prevents the installation of an unsecured experiment which could introduce, as a positive step change, sufficient reactivity to place the reactor

in a transient that would cause a violation of a limit as analyzed in Section 13.2.2 of the SAR. e. Specification 3.8.e assures that the reactivity wmth of all unsecured experiments shall not exceed the maximum value authorized for a single secured removable experiment. f. Specification 3.8.f restricts the generation of hazardous materials to levels that can be handled safely and easily. With the exception of the SGE target rods, analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR and Section 10.1 of Amendment No. X). g. Specification 3.8.g restricts the generation of hazardous materials to levels that can be handled safely and easily. Analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted (Ref. Section 13.2.6 of the SAR). h. Specification 3.8.h provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). I. Specification 3.8.i is intended to reduce the likelihood of damage to reactor or pool components resulting from the detonation of explosive materials (Ref. Section 13.2.6 of the SAR). J. Specification 3.8.j provides assurance that no chemical reaction will take place to adversely affect the reactor or its components. k. The extremely low temperatures of the cryogenic liquids present structural problems that enhance the potential of an experiment failure. Specification 3.8.k provides for the proper review of proposed experiments containing or using cryogenic materials. 1. Specification 3.8.l provides assurance that the integrity of the beamports will be maintained for all loop-type experiments. m. Specification 3.8.m assures that corrosive materials which are chemically incompatible with reactor components, highly flammable materials, and toxic materials are adequately controlled and that this information is disseminated to all reactor users. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.8 Experiments -Continued n. The limitation on experiment materials imposed by Specification 3.8.n assures that the limits of 10 CFR 20, Appendix B, are not exceeded in the event of an experiment failure. o. -p. Specifications 3.8.o and 3.8.p provide guidance for experiment safety analysis to assure that anticipated transients will not result in radioactivity release and that experiments will not jeopardize the safe operation of the reactor. q. Specification 3.8.q is intended to reduce the likelihood of accidental voiding in the reactor core or water annulus surrounding the center test hole by restricting materials which could generate or accumulate gases or vapors (Ref. Section 4.5 of the SAR). r. Specification 3.8.r is intended to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation (Ref. Sections 4.5 and 13.2.6 of the SAR and Sections 5 and 6 of Amendment No. X). s. Specification 3.8.s is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure (Ref. Section 13.2.6 of the SAR). t. Specification 3.8.t is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure. Amendment No. X provides the safety analysis which shows why the SGE target rod can be excepted from Specification 3.8.t. u. Specification 3.8.u is intended to limit the experiments that can be moved in the center test hole while the reactor is operating to those that will not introduce reactivity transients more severe than one that can be controlled without initiating reactor safety system action (Ref. Section 13.2.2 of the SAR). v. Specification 3.8.v provides assurance that the processing of iodine-131 can be performed safely and that equipment necessary for accident mitigation has been installed (Ref. Amendment No. 37). A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.11 Selective Gas Extraction Target Experimental Facility Applicability: This specification applies to the Selective Gas Extraction (SGE) Target Experimental Facility (TEF). Objective: The objective of this specification is to reasonably assure that the health and safety of the staff and public is not endangered as a result of operation of the SGE TEF. Specification: a. Safety Limits: (1) The temperature of a SGE target rod low-enriched uranium dioxide pellet shall not exceed 5180 °F (2860 °C) for any operating condition. (2) The temperature of a SGE target rod Zircaloy-4 cladding shall not exceed 1652 °F (900 °C) for any operating condition. b. Limiting Safety System Settings: Mode I Operation SGE Target Assembly Coolant Low Flow Rate 95 gpm (Minimum)(!) SGE Target Assembly Coolant High Flow Rate 129 gpm (Maximum)°) SGE Heat Exchanger Outlet Water Temperature 105 °F (Maximum) (I) Each SGE target assembly. c. Each SGE target cartridge shall contain eleven (11) SGE target rods. d. Each SGE target rod shall not be iITadiated for greater than 480 hours0.00556 days 0.133 hours 7.936508e-4 weeks 1.8264e-4 months at ten megawatts. e. The SGE TEF shall not be operated unless the following components or systems are operable: (1) One (1) SGE TEF coolant pump and one (1) SGE TEF coolant heat exchanger; and (2) SGE TEF Decay Heat Removal System. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.11 Selective Gas Extraction Target Experimental Facility -Continued f. The SGE TEF shall not be operated unless the following minimum number of radiation monitoring channels are operable. 1. Radiation Monitoring Channel Number Pool Coolant Radiation Monitor 1 (I) (I) Exception: The pool coolant system shall be sampled and analyzed at least once every four (4) hours for evidence of SGE target rod failure if the pool coolant radiation monitor is inoperable. g. The SGE TEF shall not be operated unless the following instrument channels are operable: Instrument Channel Number 1. SGE Target Assembly Inlet Water Temperature 1 2. SGE Target Assembly Outlet Water Temperature 1 Bases: a. The maximum normal operating temperature limit for the SGE target rods is established such that even at their hottest conditions of a maximum power level and a minimum flow rate loss of forced coolant accident, the maximum pellet temperature will still be below the uranium dioxide melting temperature of 5180 °F (2860 °C). The design limit on Zircaly-4 cladding is the predicted temperature for the onset of rapid oxidation of 1652 °F (900 °C). The cladding temperature predicted by transient analysis for the most severe loss of forced coolant accident is 1202 °F (650 °C), well below the onset ofrapid oxidation temperature. Further, the same analysis shows that during this time, retraction of the contained uranium dioxide pellet eliminates any compromise of the cladding mechanical strength. b. The limiting safety system settings (LSSS) for the SGE TEF are set points which, if exceeded, will cause the reactor safety system to initiate a reactor scram to prevent a safety limit for the SGE target rods from being exceeded. The LSSS help ensure that coolant flow rate and temperature will remain within the operating limits for the SGE TEF under the most severe anticipated transients. c. Thermal steady-state and transient analyses are based on the SGE target cartridge being fully loaded with eleven (11) SGE target rods (Ref. Amendment No. X). A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.11 Selective Gas Extraction Target Experimental Facility -Continued d. Thermal steady-state and transient analyses are based on an SGE target rod being irradiated for no greater than 480 hours0.00556 days 0.133 hours 7.936508e-4 weeks 1.8264e-4 months at ten megawatts (Ref. Amendment No. X). e. Specification 3.11.e assures that the SGE TEF can operate safely during state and transient conditions. f. Specification 3.11.f provides for the early detection of a leaking SGE target rod so that corrective actions can be taken to minimize the release of fission products. g. Specification 3 .11.g assures that sufficient instrumentation is available to safely operate the SGE TEF. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.11 Selective Gas Extraction Target Experimental Facility Applicability: This specification applies to the surveillance requirements of the Select Gas Extraction (SGE) Target Experimental Facility (TEF). Objective: The objective of this specification is to reasonably assure proper operation of the SGE TEF. Specification: a. Each SGE target cartridge shall be verified to consist of eleven (11) SGE target rods prior to inserting the SGE target cartridge into the SGE target assembly. b. The operating history of each SGE target rod shall be verified prior to placing it into an SGE target cartridge to ensure that the SGE target rod will not be irradiated for greater than 480 hours0.00556 days 0.133 hours 7.936508e-4 weeks 1.8264e-4 months at ten megawatts. c. The following components or systems shall be tested for operability at monthly intervals except during extended shutdown periods when the components or systems shall be tested prior to SGE TEF operation: (1) One (1) SGE TEF coolant pump and one (1) SGE TEF coolant heat exchanger; and (2) SGE TEF decay heat removal system. d. The radiation monitor as required by Specification 3 .11.f shall be channel calibrated on a semiannual basis. e. The radiation monitor as required by Specification 3 .11.f shall be checked for operability with a radiation source at monthly intervals. f. The instrumentation required to monitor the parameters required by Specification 3 .11.g shall be channel calibrated on a semiannual basis. g. A thermal power verification of SGE TEF power, using coolant flows and differential temperatures, shall be performed weekly when the reactor is operating at 10 MW. Bases: a. Specification 4.11.a assures that the SGE target cartridges have the correct number of SGE target rods as assumed in the thermal-hydraulic and transient analyses (Ref. Amendment No. X). A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.11 Selective Gas Extraction Target Experimental Facility -Continued b. Specification 4.11.b assures that the SGE target rods are not irradiated for greater than 480 hours0.00556 days 0.133 hours 7.936508e-4 weeks 1.8264e-4 months as assumed in the thermal-hydraulic and transient analyses (Ref. Amendment No. X). c. Monthly testing of the SGE TEF coolant circulation pumps, coolant heat exchangers, and decay heat removal system is adequate to provide assurance of continued operability. d. Semiannual channel calibration of the radiation monitoring instrumentation will assure that long-term drift of the channels will be corrected. e. Experience has shown that monthly verification of operability of the radiation monitoring instrumentation is adequate assurance of proper operation over a long time period. f. Semiannual channel calibration of the instrument channels will assure that term drift of the channels will be corrected. g. Thermal power verification of the SGE TEF helps ensure that the reactor state thermal power limit is not being exceeded. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 5.7 Selective Gas Extraction Target Experimental Facility Applicability: This specification applies to the design of the Selective Gas Extraction (SGE) Target Experimental Facility (TEF) and target rods. Objective: The objective of this specification is to assure compatibility of the SGE TEF and target rods with the reactor. Specification: a. The SGE TEF shall consist of not less than two (2) coolant pumps, two (2) coolant heat exchangers, plus all associated piping and valves. One (1) coolant pump and one ( 1) coolant heat exchanger shall be designed to provide the necessary cooling for either one (1) or two (2) operating target assemblies; the other coolant pump and heat exchanger are installed spares. b. The secondary coolant system shall be capable of continuous discharge of heat generated by the SGE TEF. c. The coolant pumps and heat exchangers of the SGE TEF shall constitute two (2) parallel systems separately instrumented to permit safe operation at ten megawatts, on either system operating, for either one (1) or two (2) operating target assemblies. d. The SGE TEF shall have a decay heat removal system. e. All major components of the SGE TEF in contact with pool water shall be constructed principally of aluminum alloys, stainless steel or Zircaloy-4. f. Each SGE target rod shall contain nominally 100 low-enriched uranium dioxide pellets with a nominal active length of 23.6 inches (600 mm) and a target rod nominal total length of 26.5 inches (673 mm). g. The SGE target rod low-enriched uranium dioxide pellets shall have a nominal height of 0.236 inches (6 mm) and a nominal outside diameter of 0.197 inches (5 mm). h. The SGE target rod low-enriched uranium dioxide pellets shall be nominally enriched to 19.75% in the isotope uranium-235. I. Each SGE target rod shall have a maximum uranium-235 loading of 21.6 grams. j. The SGE target rod cladding material shall be Zircaloy-4. A-X UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 5.7 Selective Gas Extraction Target Experimental Facility -Continued k. The SOE target rod cladding shall have a nominal inside diameter of 0.201 inches (5.1 mm) and a nominal outside diameter of 0.240 inches (6.1 mm). 1. The SOE target rods shall be contained in the SOE TEF target assemblies, in-pool target rod storage locations, or fresh target rod storage locations. m. All SOE target rods shall be stored in a geometrical an-ay where the value of Keff is less than 0.9 under all conditions of moderation and reflection. n. In-adiated SOE target rods shall be stored in an an-ay which will permit sufficient natural convection cooling such that the temperature of the target rods will not exceed design values. Bases: a. -e. The SOE TEF is described and analyzed in Amendment No. X. The SOE TEF and target rods can be safely operated at a reactor power level often megawatts as described in the Amendment. f. -k. Specifications 5.7.f, 5.7.g, 5.7.h, 5.7.i, 5.7.j and 5.3.k require uranium content, materials and dimensions of the SOE target rods to be in accordance with the design and fabrication specifications (Amendment No. X). 1. -n. The limits imposed by Specifications 5.7.1, 5.7.m and 5.7.n are conservative and assure safe SOE target rod storage. A-X ATTACHMENT 3 CODES AND STANDARDS 1. Design, Fabrication and Operation General Atomics (GA) is the prime contractor for the design and supply of the Selective Gas Extraction (SGE) Target Experimental Facility (TEF) structures, systems and components (SSC). The SGE TEF was designed and fabricated in accordance with applicable codes and standards, specifically:

The SGE TEF SSC and conduct of operations shall comply with all applicable U.S. Nuclear Regulatory Commission (NRC) requirements, other federal regulatory requirements, as well as local and state design codes and standards requirements.

SGE TEF components shall be designed to meet applicable requirements in Section VIII, Division 1 and 2 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 2015 (the term applicable refers to the method of evaluating the condition of structures and limits to stress, strains and cumulative fatigue damage).

Welding of the in-pool target and cooling systems shall meet the applicable requirements1 in Section IX of the ASME B&PV Code, 2015. Weld nondestructive examination (NDE) shall meet the intent of the applicable requirements in Section V of the ASME B&PV Code, 2015.

All materials used in the SGE TEF components shall meet the intent of applicable requirements in Section II of the ASME B&PV Code, 2015. 2. Design Methods 2.1 Software The nuclear, thennal-hydraulic, and mechanical design of the SGE TEF utilized a variety of analytical and design software packages that include:

Monte Carlo code MCNP6 [LANL 2003] using ENDF/B-VIl.1 for the nuclear design of the target assembly.

RELAP5 and FRAPTRAN to confirm thermal-hydraulic parameters of the target assembly, and for analysis of SGE TEF transient conditions.

MCODE, an open source linkage-program for combining MCNP6 and ORIGEN2.

ORIGEN2 and MATLAB for radionuclide source term calculations.

FRAPCON for target rod performance verification including pellet-clad mechanical interaction (PCMI), pellet-clad chemical interaction (PCCI) and internal pressure buildup from volatile radionuclide release from the pellets.

ANSYS 2016 for the structural performance analysis for the target assembly. The applicable SSCs of the SGE facility will be designed to meet all the requirements of the ASME code, but will not be code certified SSCs. 1of5 ATTACHMENT 3

ANSYS-FLUENT for single and two-phase flow modeling in the target cartridge during normal operation.

SINDA for modeling thennal-hydraulic perfonnance of the target cartridge.

Mathcad for SSC sizing analysis.

SolidW orks 2016 for the detailed mechanical design and 3D visualization.

AutoPIPE for structural analysis of the Target Cooling System. The use of all software for the development of the SGE TEF has been subjected to the rigorous software quality assurance (QA) verification and validation procedures as required by the GA QA Program and subtier documents (Section 3), including preparation of verification and validation reports as required by the applicable requirements in the engineering procedures (Ref. 1 ). 2.2 Design Documentation Design bases, including all assumptions for the design bases, as well as documentation of software verification calculations that fonn the bases of the design and safety information presented in the license amendment are provided in the following reports listed in Table 1. 2 This table also includes other applicable documents. Table 1 Design and Software Validation Reports and Other Applicable Documents REPORT NO. DESIGN CALCULATION REPORT TITLE 30441R00017 ANSYS Target Cartridge, Housing Structural Analysis Design Calculation Report 30441R00019 Target System Cooling Calculation Report 30441R00021 Target Assembly Thermal Analysis 30441R00022 Source Term Analysis Design Calculation Report 30441R00030 Mo-99 Target Cooling System Seismic Analysis Design Calculation Report 30441R00031 Mo-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00032 RELAP Accident Analysis and FRAPTRAN Target Rod Transient Analysis Design Calculation Report 30441R00033 Analysis of Forced Convection Cooling of Target Rods with 2 Phase Considerations 30441R00035 Cooling of the MURR Beryllium Reflector 30441R00038 Computational Fluid Dynamics Analysis of Target Housing Design Calculation Report 2 All reports listed in Table I are considered GA Proprietary Information and will be appropriately marked. 2 of5 ATTACHMENT 3 Table 1 (con't) Design and Software Validation Reports and Other Applicable Documents REPORT NO. DESIGN CALCULATION REPORT TITLE 30441M00043 ANSYS Thermal Model ofRB-MSS Target Rod REPORT NO. SOFTWARE VERIFICATION TEST 30441R00002 MCNP6 (Version 1.0) Verification 30441R00003 ORIGEN2 (Version 2.2) Verification 30441R00006 MCODE (Version 2.4) Verification 30441R00020 ANSYS Workbench (Version 16.0) Verification 30441R00023 OTSTM (Version 1.0) Verification 30441R00024 RELAP (Version 5 Mod. 3.3 P03) Verification 30441R00025 FRAPCON (Version 4.0) Verification 30441R00026 FRAPTRAN (Version 2.0) Verification 30441R00028 ANSYS-FLUENT (Version 16.0) Verification 30441R00029 SINDA (Version 2.2) Verification 30441R00034 OTSTM (Version 1.0) Software Design Requirements, Description and Verification REPORT NO. OTHER APPLICABLE DOCUMENTS 30441S00001 Molybdenum-99 Supply System Requirements Document QAPD-30441-11 Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System (RB-MSS) 3. Quality Assurance GA conducted the analysis and design, and will fabricate the required SSCs under a QA program that ensures conformance, through a documented system of QA requirements, specifications and inspections, with the applicable requirements of ASME NQA-1 2008/-la-2009 Addenda that are contained in the GA Quality Assurance Manual (QAM) (Ref. 2) and the accompanying Quality Division Instructions (QDI) (Ref. 3). GA, as the prime contractor for the supply of the SGE TEF SSCs, will follow a graded approach in the implementation of its QA program commensurate with the Quality Assurance Levels (QAL) requirements on both safety and performance of the facility. The QA requirements from GA's QA program, and the customer requirements, are implemented via project specific Quality Assurance Project Documents (QAPD) which establishes GA's QA requirements applicable to SGE project specific design, analysis, and fabrication related activities (Refs. 1 and 4). 3of5 ATTACHMENT 3 3.1 Quality Assurance Level of Safety-Related Components Project/Resource Procedures Manual, procedure EP-4010 (Ref. 5), discusses the determination and assignment of QA levels. Three characteristics make SSCs safety-related. If an SSC has one or more of these characteristics, then it is considered to be safety-related. 1. Maintains the integrity of the reactor coolant boundary (the reactor vessel and associated piping that circulates the reactor coolant). 2. Has the capability to shut down the reactor and maintain it in a safe condition. 3. Has the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures. EP-4010 also states that components which contain radioactive materials are classified as QAL I as well as SSCs designated engineered safety features. 3.1.1 Selective Gas Extraction Target Experimental Facility QAL 1 Components As the result of the engineering determination and assessment to identify all of the equipment required for the SGE TEF, the following components are considered to meet the QAL I designation:

The target rods' cladding, endcaps (upper and lower) and welding of the endcaps.

The target pellets.

The flow meters in the target cooling system.

Temperature sensor downstream of the heat exchanger in the target cooling system, including its signal conditioner. 3.2 Manufacture of Low-Enriched Uranium Target Rods In addition to following the graded approach for SSCs under the QAL approach mandated by GA's NQA-1 compliant QA program and implemented via the project specific QAPDs, the fabrication of enriched uranium (LEU) target rods will be subjected to a the following Quality Control procedures: 3.2.1 Target Pellet Fabrication

Qualification of the manufacturing process.

Ensuring traceability of pellet lots through a system of fabrication travelers.

Verification of the LEU pellet loadings in each pellet lot using statistical sampling.

Measurement of pellet diameter using statistical sampling. 4 of5 1 I . I ATTACHMENT 3 3.2.2 Target Rod Fabrication

Qualification of the manufacturing process.

Ensuring complete traceability of LEU pellet loadings into target rods through a system of fabrication travelers, including traceability on all cladding components.

100% leak testing on target rod welding.

Physical measurements such as target rod length. 4. References 1. Project/Resource Procedures Manual (P/RPM), General Atomics Document GA-A15466, Second Ed., February 2015, containing Engineering Procedures EP-4020, "Design Control System Description," and EP-4070, Issue C, "Control of Scientific and Engineering Computer Programs." General Atomics Proprietary Information 2. General Atomics Quality Assurance Manual, 4th Edition, 26 August 2016. 3. General Atomics Quality Division Instructions Manual, as revised, 03 June 2016. 4. Quality Assurance Program Document -Phase II, Reactor-Based Molybdenum 99 Supply System, General Atomics Document QAPD-30441-II (January, 2017) and as revised. General Atomics Proprietary Information 5. Project/Resource Procedures Manual (P/RPM), General Atomics Document GA A15466, Second Ed., February 2015, containing Engineering Procedure EP-4010, Issue A, "Safety and Quality Assurance Classifications." General Atomics Proprietary Information 5 of5

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