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Reed Response to Request for Additional Information with a Revised Safety Analysis Report and Technical Specifications

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Reed Response to Request for Additional Information with a Revised Safety Analysis Report and Technical Specifications REED RESEARCH REACOR LICENSE NO. R-112 DOCKET NO. 50-288 REED RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION WITH A REVISED SAFETY ANALYSIS REPORT AND TECHNICAL SPECIFICATIONS JULY 2010 REDACTED VERSION* SECURITY-RELATED INFORMATION REMOVED *REDACTED TEXT AND FIGURES BLACKED OUR OR DENOTED BY BRACKETS REED COLLEGE REACTOR FACILITY 3203 Southeast Woodstock Boulevard July 30, 2010 Portland, Oregon ATTN: Document Control Desk 97202-8199 U S. Nuclear Regulatory Commission Washington, DC 20555-0001 telephonw 5031777-7222 Docket: 50-288 fax License No: R-112 503/777-7274 Subject: RAI TACNO. ME1583 email [email protected] Attached are some of the answers to the subject RAI dated March 8, 2010. web As noted under the individual RAIs, some of the information needed for hrrp://reactor.reed.edu response is to be completed as part of the analysis being performed by Oregon State and General Atomics and will be submitted by November 2010. The response and attachments do not contain any sensitive informa- tion. Please contact us if you have any questions. Thank you. I declare under penalty of perjury that the foregoing is true and correct. Executed on 7-34- '- pDirector, Reed Research Reactor. Attachments: 1. RRRRAIs 1-58 2. RRR SOP 34, Control Rods 3. RRR SOP 34B Control Rod Calibration January 2010 4. RRR SAR Chapaters that have changed since the 2007 submital Aco'-Q) pvac 1. NUREG-1537, Part 1, Section 1.4, Shared Facilitiesand Equipment, states the applicant should consider whether the loss of any shared facilities or equipment could lead to a loss of function that would lead to an uncontrolled release of radioactive material,or if released,are analyzed and found to be acceptable. The 2007 SAR, Section 1.4 discusses this subject. However, the discussion is incomplete in that it does not include the loss of electricity and how it would affect the release of radiation should it coincide with the loss of fuel cladding integrity. Please provide this information including the loss of alarms, automatic isolation, operation of heating, ventilation, and air conditioning (HVAC) systems, etc. Please provide information concerning whether the analysisprovided in Chapter 13 envelopes this condition. See updated SAR Section 1.4. Although Chapter 13 will not be completed until November 2010, that analysis will envelope this condition. 2, NUREG-1537, Part 1, Section 1.5, Comparison With Similar Facilitiesstates the applicantshould use pertinent information from other reactors and this information can be used to compare the safety envelope of Reed Research Reactor (RRR) and to support analysis in appropriatechapters of the SAR. The 2007 SAR discusses this, but the information is incomplete. Please provide a comparison of the RRR to other TRIGA facilities so as to characterize the degree to which generic information or operational experience from other reactorfacilities is applicable. See updated SAR Section 1.5. 3. NUREG-1537, Part 1, Section 2.2, Nearby Industrial, Transportation,and Military Facilities,states information on nearby military facilities be included in the SAR. The2007 SAR, Section 2.2 discusses industrialand transportationfacilities but does not discuss military installations.Please provide information concerning the nearby military installations. See updated SAR Section 2.2. 4. NUREG-1537, Part 1, Section 3.1, Design Criteria,states the applicantshould identify the design criteria that are applicable to each structure,system and component that performs.a safety function. The 2007 SAR, Section 3.1 briefly addresses this matter and states "the originalreactor installation in 1968 used fuel and components manufactured by GeneralAtomics (GA), and the specifications to-which structures were built were those stated by GA. Specific design criteria were not stated. All building modifications and equipment additions were in conformance with the building codes in existence at the time." Please provide the criteria applicable to the original design and construction and to subsequent modifications to the design and construction. See updated SAR Section 3. 1. 5. NUREG-1537, Part 1, Section 3.3, Water Damage, requires the applicantidentify the potential for flooding which could prevent structures,systems and components from performing their safety function. The 2007 SAR, Section 3.3 states "As discussed, in * Chapter 2, the flood plain of the local rivers does not come near the reactorsite. However, even if flooding occurred, reactorsafety would not be an issue since the core is located in a waterpool." However, this information is incomplete. Please provide Attachment to RRR RAI Response 20100730 I of 17 informatiori thatdemonstratesthat shouldiflooding occur, it will not prevent operation of the RRR safety systems',. See updated SAR Section 3.3. 6.. NUREG-1537, Part 1, Section 3.5,pSystems and Components, states the applicant should identify the bases or design features of the electromechanicalsystems that are used to ensure safe operation and shutdown. of the reactorduring all conditions.- ,The2007 SAR, Sections 3.5 and 3.6 provide -some:infornationon this topic but do not provide information on the design features of the control rods (e.g., fail safe in the event of loss of power) or the systems associated with reactoroperation and safety (e.g., power level scrams, interlocks to limit reactivity insertion). Please provide the design for electromechanicalsystems and components requiredifor operation, shutdown and to maintain shutdown , . See updatedSAR Section 3.5 and 3.6., 7. NUREG-1537, Part,1, Section,4.2. 1,; Reactor Fuel,,states the applicant-shoulddescribe the fuel elements used in the reactorincluding.detaileddesign information. References should be provided to demonstrate that the design basis assures-thatintegrity of the fuel is maintained under all conditions assumed in the safety analysis. The description should also,include information.necessaryto establish limiting,conditions beyond which fuel integrity would be lost. The 2007 SAR, Sections ;1.3.3, 4.2.4 and 4.2.5 which provide some information are incomplete. Please-discuss the -differences!in fuel length for the aluminum and stainless steel clad fuel utilized in the core and the implications of these differences on analysis. In addition;please address mechanicalforces and stresses, corrosion and erosion of cladding,.hydraulic,forces,,, thermal changes and temperature gradientsi.and internal pressures.from fission products and the production of fission gas. Include in the analyses the impact of radiation effects, including the maximum fission densities and fission rates that the;fuel is designed to accommodate. This will be completed as part of the analysis being performed by Oregon State and General Atomics and will be submitted by November 2010. 8. NUREG- 1537, Part 2, Section 4.2.2 states the control rods should be sufficient in. nu~mber and reactivity worth, to comply withthpe 'single stuck rod' criterion; that is, it should be possible to, shut down the reactorand comply-with the requirementof.. minimum shutdown margin with the highest worth scrammable control rod stuck-out of the core. The. control rods should also-be sufficient to control.the reactor in all.designed operating modes and to shut down the reactorsafely, from any operationalcondition. The control rods, blades, followers (if used), and support systems should be designed conservatively to withstand af/anticipated stresses and challenges from mechanical, hydraulic, and thermal forces and the,effects of their chemical and radiationenvironment. • The control rods should be designedso that scramming them does not challenge their integrity,or operationor the integrity oroperation of other reactorsystems. The 2007 SAR, Sections 4.2.7, 4.2.8 and 4.2.9, while providing some of this infonrnation, is incomplete. The SAR d6es not provide the worths of the 3 RRR control rods. Please ,,provide calculatedand measured control rod worths underall conditions bf6operation. Please determine if control rod withdrawal insertion limitation limits (rodposition vs. jioweýr) are necessary to preserve assumptions in' the'departurefrom nucleate boiling ratio (DNBR) analysis. ..... ... .. Attachment to RRR RAI Response 20100730 2 of 17 This will be completed as part,of the analysis, being, performed by Oregon State and General Atomics and will be submitted.'by November.2010. ,, 9. NUREG-1537, Part 1, Section 4.2.3, Neutron:ModeratorandReflector, states the applicantshould describe reflectors and moderators designed into the core and theirspecial features. The 2007 SAR, Sections 4.2.2 and 4.2. 6;RRR provides a discussion`ofthe radial reflectorand the-graphite reflectorelements;, however, it.does not provide any information pertaining to the naturally circulating water whichis also moderator/reflector.,Please provide a description of the watermoderator and reflectorand an assessment of the function and importance of the moderatorand,the effect of loss of moderatoron. the behaviorof the ,,reactorcore during.operations.., .: . :: ' t This will be completed as, part of the analysis being performed by Oregon State and General Atomics and will besubmitted by7NoVember 2010 , . 10. NUREG-1537, Part 1, Section 4.2.4 Neutron Stdrtup.Source, states the applicantshould describe the neutron source used for reactorstartup. The 2007 SAR, Section 4.2.10 ,protuidesa description of the neutron source holder only. Please re vieW the;cited . requirement and then supply a revised desdription of the neutron sour-ce.in use at RRR including the fcilowing; ,.
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