View source for RCF 08-04, Response to Request for Additional Information Regarding Rensselaer Polytechnic Institute Renewal Application

{{Adams
| number = ML082190523
| issue date = 07/28/2008
| title = Response to Request for Additional Information Regarding Rensselaer Polytechnic Institute Renewal Application
| author name = Winters G
| author affiliation = Rensselaer Polytechnic Institute
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000225
| license number = CX-22
| contact person = 
| case reference number = RCF 08-04
| document type = Letter type:RCF
| page count = 81
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:Rensselaer                                              DEPARTMENT OF MECHANICAL, AEROSPACE, AND NUCLEAR ENGINEERING RCF 08-04 July 28, 2008 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Re: Response to Request for Additional Information

==Dear Sir:==

This letter provides the information requested by your letter of March 21, 2008, "Request for Additional Information Regarding the Rensselaer Polytechnic Institute (RPI) Application for Renewal of Facility License No. CX-22 for the Rensselaer Polytechnic Institute Reactor Critical Facility".
The requested information is attached in several parts. Specifically:
Responses to the detailed questions Revised Chapter 12, Safety Analysis Report Revised Technical Specifications Explanation of Changes to the Technical Specifications The Nuclear Safety Review Board has reviewed and approves the revised Technical Specifications.
Sincerely; Glenn Winters, Director L. David Walthousen Laboratory I declare under penalty of perjury that the foregoing is true and correct.
Executed on:
Date                         Dr. Timothy Wei, Iherim Dean of Engineering Rensselaer Polytechnic Institute 110 8th Street I Troy, NY 12180-3590 USA

cc:
Dr. Michael Podowski, Chair       Dr. Jeffery Geuther, Operations Supervisor RPI NSRB                          L. David Walthousen Laboratory Dr. Achille Messac, Interim Chair Dr. Peter Caracappa MANE                              Radiation Safety Officer Dr. Timothy Wei, Interim Dean     Dr. Timothy Trumbull, Adjunct Professor School of Engineering             MANE William Kennedy, NRC              Glenn Winters, Director L David Walthousen Laboratoy

RAI Response RPI Responses to RAI dated March 21, 2008 1.1 Section 1.3, General Description of the Facility. Figure 1.1 shows an overhead trolley-type crane in the Reactor Room. However, no description is included in the Relicensing Report. Discussions with facility personnel indicate that this crane has not been used for some time, and that power to it is removed at the breaker. In the event this crane is used in the future, describe what inspections will be conducted and what controls will be put in place for movement of loads over vital equipment.
RPI Response - The crane in the Reactor Room is operable and will be inspected periodically by a subcontractor to ANSI and OSHA standards. This is planned to commence in Fall 2008. The most recent inspection was 2004. The only restrictions judged necessary are to restrict lifting loads over the reactor tank when the reactor is fueled. No other lifting casualty would endanger equipment critical to reactor safety.
Crane operation is under the supervision of the duty Senior Reactor Operator who is responsible to enforce the necessary restrictions.
3.1 - Sections 3.3, Water Damage, and 2.4, Hydrology. The SAR indicates that the Mohawk River flood stage has exceeded the elevation of the reactor room floor in the last century and has repeatedly exceeded that of the reactor water storage tank pit floor.
When considering ground water entering the reactor water storage tank pit or river water entering the reactor room, address the probability of occurrence, safety consequences, access to the building during flooding, and contingency plans that are in place if needed.
The issue of concern is that the fuel is maintained in a subcritical configuration and is physically secure.
RPI Response - The facility ground floor is 2 feet above the 100-year flood level, but' below the 500-year flood level. Facility flooding would potentially damage equipment, but the normal secured shutdown status requires no operable equipment to ensure reactor safety. The fuel stored in the vault is subcritical when fully flooded (see the response to RAI 9.1), as would be the fueled core if fully flooded. There are no plans to attempt to remove the reactor fuel when a flood is pending since it is judged that the removal process would be more hazardous than leaving the fuel in the safe configuration within the facility. The facility would be guarded whenever it was vulnerable, such as if the intrusion detection system was inoperable, to prevent unauthorized access. Further, the absence of radioactive contamination, verified by periodic sampling, means that flooding would not release any radioactive material.
4.1 - Section 4, Reactor Description. Include a discussion of the auxiliary reactor scram (moderator-reflector water dump) in Chapter 4. What are the criteria for when the moderator dump feature is required and when it can be bypassed?
RPI Response - The following explanation of the auxiliary reactor scram will be added to the SAR. The addition will be a paragraph at the end of Section 4.3:
 "The reactor tank is filled by pumping water from the storage tank with the Reactor Fill Pump. Water may be drained slowly from the tank through the fill line by operating the air-operated drain valve. See Figure 5.1. Water may be quickly drained from the reactor Page 1

RAI Response tank by,0peningthe'6-inch, butterfly valve~on the fast dump line. This valve is also air-operated and fails. open on loss, of air pressure or loss of site &#xfd;power.:When the dump valve is opened water rapidly drains backs into the storage .tank and ensures, thereactor is shutdown regardless of the position of the control rods. The dump valve automatically opens if a reactor scram occurs. A kebd         ,swiich     on the Auxiliary Control Panel, CP-2, can override this' automatic trip when .lOwed by opeatfing procedures."
Operating procedures allow the, automatic tip, of the dump valve to,be bypassed if the duty Senior Reactor Operator concurs, and if the .reactor contains a. known core. A known core is one for which several core parameters hgvebeenr measured to yerify the core configuration is within the design envelope analyzed for casualties. A specific definition of a known core is contained, in the Technical Specifications.
4.2.- Section 4.2.1,; Reactor FueL. 'This se'ction* of tihe SAR statdS thaith'
                             . -i,...                                           teSPERT (F-i) :
4 fuel pin design 'was previously qualified by the DOE 'ahd' NRC: NUREG-1281, "Evaluation. of the Qualification of SPERT Fuel for Use in Non-Power Reactors," August I987; is the report on the NRC's evaluatiion of the qualification of your fuel. Provide any inf6rmation that may hav';e abeainng on the conclusions of NJURG-1281' Or the suitability ofyour fuel duiig the period of the renewed icense:
RPI Response - The SPERT (F-i) 'fuel tins provided !to RPI Were determined to be suitablefor low power *use 6y  5 the'e~ial'ation&#xfd;       reported in'NUREG-1281. The conclusion remains valid today' stnce the fuei has co'ntin uiedto be useid:in the lovt power RPI reactor.
The fuel is stored dry betwedn re'ator operations2 The typical reactor operation consists of a few minutes at'powerlevels*belof I: i00:,watts. After operation, the moderator is.
drained from the reactof tan. Thd fu'6l pin's areiett d foi' only'a few houts considering' time to prepare to operate, opjefating time;'and timid to secure the facility to secure'the facility. Visual inspection and contamination surveys have not shown any corroded or oth'erwise defectve pins.,Surveys of fuel pin sho no sxgnificant buildup of fission products. Radiation measu'rerments a few' days ater reactor operation show radiation levels of about 1 millirem per hour at contact on a typical fuel pin.
4.3 - Section 4.2.2, Control Rods. TheSSAR stateg that' tie 66dn*ol rod dri;ves are designed so rods can be located anywhere in the tank. Clarify whether the intended license basis is restricted to the core arrangemnent described in the SAR oi'assufmes the use of other control rod configurations. If thelatte', provide additional' discussion on the design boundaries, safety review process and acceptance criteria fdr core redesigns.
Considering that information, propose Technical Specifications ,(TS) that ensure configuration control    .. ,,                     ...                                 ,.
RPI Response - The, controlrod ganties-can'be' swiveled.and extended to change the position of the control rods in the tank. Moving the control rods would require cuts to be made in the lattice plates, and they have not been moved since the core was reLfueled with LEU. Reactor physical configuration is intentionally flexible to allow experiments with varying configurations..
Page 2

RAI Response The design boundaries are included in'the Techinical Specifications. These boundaries specify the, number of control rods'(four) and shutdown reactivity. Planned deviation from these constraints would require a license revisio '".
Part of the qualification process of a core given in the technical specifications is the measure of control roq worth to verify that rod worth, drop time, and shutdown margin requirements are met. The critical loading procedure for an unknown core is a very conservative app'roach requiring that the reator'moderator is drained after every group of pins is loaded (using the inversermiultiplicatio'n approach) and that the excess reactivity is measured after each group'of pins'i loaded.
4.4 Section 4.2.2, Control Rods (or Section 4.5.2, Reactor Core Physics Parameters). If the control rods can be withdrawn as, a gang, verify that the maximum rate of reactivity insertion due to gang control rod withdrawal. is bounded by the requirements of TS 3.2.3.
RPI Response - The bank control rod worth is approximately $2.00 over a 36" stroke, for an average differential worth of less. than,6 wits per inch. The rod bank withdrawal rate is 3 inches per minute. Therefore, the average reactivity insertion rate is approximately 6 cents / inch
* 3 inches / minute
* 1 minute / 60 seconds = 0.3 cents ($0.003) per second.
The technical specifications currently require that the maximum reactivity insertion rate due to bank withdrawal be fiye .ents ($.05) per second when the flux is greater than ten times source level. This would require the' maximum differential bank worth to be at.
least sixteen times greater than the average Worth,, which is not credible given the sinusoidal shape of the differential jank worth,curye. The revised Technical Specification no longer has a limit on control rod worth since control rod differential worth does not have. a role in the accident anlysis in the SAR.
4.5 Section 4.2.4, Neutron Startup Source. Describe the personnel shielding that exists as the neutron source is being withdrawn fro the core into,the paraffin shield.
RPI Response - The shielding present consists of the physical structure and the accessibility limitation imposed by it. Typical exposure rates at the nearest accessible points.are:      .
Source stowed in paraffin shield: 12 nmrem/hr.
Source exposed, tank empty: 23 mreI.hr Source exposed, tank full:. 2.3 mrem/hr,.
4.6 Sections 4.2.5, Core Support Structure, and 4.3, Reactor Tank. Discuss any age-related degradation of the core support structure, the reactor tank, and piping.
Discuss' any inspections that have been performed-on~such structures and'systems, the results, and any planned actions, to correct or manage age~related degradation.
RPI Response - The core support structure, piping, and reactor tank are not subject to high temperature or pressure. Therefore, age-related degradation is not expected. The core moderator is not radioactive, and is not heated, such that there is no major negative Page 3

RAI Response consequence of a hypothetical moderatorleak (see responseto 4.7'below). Due to the low probability of age-related degradation and thd low impact of the consequences of a, moderator leak, no:inspections are p effornmed to verify the integrity of these items.
4.7 Section 4.3, Reactor Tank. Discusgsthrelikelihood and 'consequences of leaks. In the event of a coolant leak from the reactor tank, the storage 'tank; or the associated piping, whatfprovisions, if any, are there to contain the Ieak and prevent!an uncontrolled release to unrestricted areas, including groundlwater? Is the coolant analyzed periodically for radioactiyity so that an estimate of any release can be documented? (This ques~tion is related to compliance with 10 CFR20.1501.)
RPI Response - The reactor moderator is not monitored on'a periodic bagis, but'is always monitored prior to release. The sensitivity of the procedure to measure;gross alpha/beta is on the order of a few: dpm perliter, and has not exceeded-the minimum detectable activity in memory. This is primarily due'fo the low operating power level: Therefore, any leak of reactor moderator, even if., it wereto reach groundwater, would not have a measureable radiologicalirmpact:      "     "        ,       .
4.8 Section 4.4, Biological Shielding. This section of th6 SAR indicates' that'the shielding is adequate for the power of 1 Watt' Please indicate typical radiation" levels to show that there is adequate shielding at the!libPdnsed power of 100 watts.' Describe controls used to ensure ALARA:during 0operati0n (e.'g.;'ro-of access control duriing operation),
RPI Response - Based upon radiation measurements at approximately 13 W, the maximum anticipated dose at-the fence-line boundary at full power (100,W), operation would be 1.3 mrem/hr, which complies with the liniitaftiodof nob more than 2trem/hr 2'         to areas accessible to members: of the public.            .
The' highest dose rate at an accessiblc 10cation inside the reactor building (just outside the reactor room door) at 100 W is 27 'hirmlhe     i (sum gamm.rna and neution).          '
The revised Technical Specification establishes an annual integrated po-wer'lim-it of-2 kWh, which Would limit the maximum possible dose inr restricted areas to&#xfd;less than       '
5 rem/year and in unrestricted areas' to less' thath-100 torem/year without the need for any further area or .access controls during operation.,
No other functions, suoh as buildings or grounds maintenance, are performedat the facility during reactoroperation. Personnel enter the reactor room area during operation only for specifically approved procedures, and any new procedures would need to be, reviewed for ALARA considerations before being implemented.
4.9 Section 4.5.2, Reactor Core Physics Parameters. Section 4.5.2 does'not list any core physics parameters. Temperature and void coefficients are found in Tables 13.2 and Page 4

RAI Response 13.3. Shutdown margin i's only given as a lower bound (> 0.02)rin Table 13.2. Please provide quantitative values for excess reactivity, and:shutdown margin in Chapter 4 and ensure that these values are consistent with the technical specifications. (See SAR RAI 13.5, TS RAI 1.3.V, and TS RAI 3.2 (D))
RPI Response - . Sh~utdown reactivity measurements are completed as part of the process of qualifying a known core to yerify that the shutdown reactivity meets technical specifications. A typical known core has 10 -. 30 cents of excess reactivity from control rod travel beyond the critical pdsifior*. The limit is 60 cents and includes the reactivity change that may be caused by a movable experimient if one is installed* The control rod bank worth is approximately.$2. Therefore, kndwir cores meet the requirement of $1 shutdown reactivity with the full foui~rod bank arnd $0.70 with a stuck control rod, given in TS 1.3 and TS 3.2, respectively. The SAR will be revised wit the next update. A revised Table 13.2is provided with the response~to RAI 13,5 below.
: 5. REACTOR COOLANT SYSTEMS 5.1 Discuss water quality requirements and the process used to maintain; water quality to minimize, corrosion and to assure adequate, visibility to -safely handle fuel elements.
RPI Response -The water used as moderator is, Schenectady city water. No chemistry controls are used since the wetted portions of the reactor water systems are stainless steel and the storage tank and piping to the fill pump are the only portions of the system that are continuously immersed. The ,reactor tankis! normally dry with moderator present for just a few hours each time the reactor is operated. Asmall pump with a wound-cotton filter is routinely run on circulation with storage.tank water to maintain water clarity.. If clarity becomes a problem, the storagetankrmay be sampled to verify no detectable activity and discharged. In practice, tank discharges are infrequent. The most recent, discharge was November 2006 in order to prepare for replacement of the fill pump.
5.2 Discuss the allowable range of-reactor tahk'Waterlevel for reactor operation and the''
technical basis. (See TS RAI 34:2.6 (B)).
RPI Response - Operating procedures require water level to be high enough to immerse the.control rod buffer pistons on the lower support plate. These pistons actas hydraulic shock absorbers at the end of the control rod stroke. The corresponding reactor tank water level is 19.5 inches. At the upper end, water level is not allowed to be more than 10 inches abovethe top of the core, This corresponds to 68 inches of water in the reactor tank. The basis for this upper limit is to provide an adequate upperreflector, notflood out instrument dry wells, and minimize the time for fast dump to reduce core reactivity.,
Most operations are measurements of reactor parameters when thecore is fully flooded, that is, water level is 10 inches above the top of the upper support plate. Some measurements are made with water'levels below the top of the upper suppoi- plate. These are benchmark measurements to determine'the critical water level, that,is, the water level for exactly critical when control rods are fully withdrawn.
Page 5

RAI Response 5.3. Describe operating procedures, interlocks; :alarms, and administrative controls that.
exist to controlthe water level in the reactor tank and .to assure that there&#xfd; is sufficient free volume in the reactor water storage tank for a reactor,tank dump in the event of a scram.
RPI Response - The Startup Checklist forbids control rod moyement unless reactor tank water level is-above the carrier plate that is, at least 19.-5 inches. There is no interlock to prevent   rod motion at lower water levels. The sanie Startup Checklist recI uires' filling the reactor tank to 68 inches' Again there is' no interlock associated with this level and there critical water is no alarm. Operation at some other water level, for example, to determine level, is done using an experiment procedure for the specific measurement being made.
TheStorage Tank is the normal repository for!the moderator at.the, faoility.;When the reactor is prepared for operation, moderator is pumped fromh the Storage' Tank into the Reactor Tank. Thus, there is .always free volume in the Storage Tank for-the Water in the Reactor Tank..'
5.4 Discuss the maximum potential level of contamination that could exist in water that collects in the sump and the likelihood and consequence of release to the environment through cracks in the concrete. (See SAR RAI 4.7),
RPI Response - As stated in the response to SAR RAI 4.7, the, maximum activity in moderator water (giveti,the detection li'mits of the available equipment) is a few dpm/L, which would result in miifiimal impacts from releaseto the environment.
7.1 Section 7.1, Summary Description. The versio4,,of the SCAR currently under review was submitted to the NRC in November 2002. As discussed in Section 7.1, substantial instrument and control (I&C) equipment upgrades were in progress at that time. To facilitate the current review please provide the,follpowing information:
: a. A more detailed description of the ob i ctive, s 66       dsign, andcurn status of instrument system upgrade project.
RPI Response - Instrumentation upgrades are complete. The following equipfent was replaced with equipment of similar capability:
Picoammeters for,three uncompensated ion chambersmeasure and-displiay current and have output signals to recorders. These repladei similar, equipment for instrumnent channels PP2; LP 1 andLP2.-The:replacement equipment was custom manufactured by Circuit Equipment Corporation .to RPI specifications.
paper strip chart! recorders were replaced with videographic recorders. Three of the videographic iecorders are manufact6red by Thermo Westrofiics. The fourth videographic-recorder is manufactuedby Honeywell. All recorders are commercially available equipment. One recorder contains ala'rm relays that are used to imple'ment the control rod outmotion interlock on low startup channel counts and loss of recorder power.
The remaining recorders do not provide any control function other than displaying moderator temperature and reactor neutron level as determined from the uncompensated ion chambers and the startup channel BF3 detectors.,               ..                          "
Page 6

RAI Response The&#xfd; electronics to receive and processTsignals-from the s.BF3, detectors were replaced with like-kind, commercially available equipment manufactured by Ortec and. Canberra.T These replacements include: high voltage power supplies, counter/scalers; preamplifiers, and discriminators.
objective of the The *~~~~~~~         I instrument
                         &#xfd; - ,*
upgradei was
:,.''*, 1. 1 to replace
                                                        .'i &#xfd;;; , '  :**,
aging and i_
unreliable equipment with commercially aviaidable, or with custom-built hardware in cases where commercially available equipipent was inadequate. Another objective was to retain all prior control and interlock functions with the new equipment. Both 'objectives were met.
A water level detector was- also procured,' bit is not yet installed. The intent of this equipmentis to. display reactor tank.water 'leVel, on a recorder. ,The future plan for the water level detector is to. interlock the reactor fill:pump with a high level shutoff and; interlockthe reactor tank'immersion heaters;,with a low Jlevel shutoff. It is also. possible that a low water level rod block could be implemented. None of these additional controls based on tank water level affect the severity or likelihood of any plant casualty.
: b. Provide enough information such that the staff can eaiuate the acceptability of the instrumentation and control presently installed. If this involves digital' equipment consider NRC Regulatory Issue.Summary 2002-22, "Use of EPRI/NEI
        'JointTask Force Rei*0rt, Guidelinie .on Licensing Digital Upgrades: EPRI TR-'
102348, Revision 1, NEI',01 A'l:                  Revision ofnEPRI TR-102348 To Reflect Changes to the 10 CFR 50.59 Ridle."
RPI Response'- DetAils of the cnurrently ihstalied i'nstuntientation are provided in the response td question 7.2e       bo*w.
7.2 Section 7.2.1, Design Criteria. The iifnfdrmation presented in Sectioha 71.2.1 is limited to a brief, general, description of the functions of the I&C systems. Expand this section to'describe the criteria (standards, codes, and gidehnes) that form the design bases of the I&C systems (

==Reference:==
NUREG. 1537 Part I, Format and Content Guide, Section 7.2)..                        .'
RPIResponse'-.The reactor operating status is.monitored by two types of neutron.
detectors.-, BF3, detectors are the more sensitive and show changes in reactor neutron level when the reactoris shutdown or beginning:a startup. Two such' detectors'ate installed and are labeled Startup channels A and B. Uncompensated ion chambers areless sensitive and are used as the reactor approaches operating power levels. Three such detectors are..
installed. Overlap'is achieved by the differinig detector'sensitivities and by the l6cation of the individual detectors. The arrangement ensures that two or more neutron detectors are always able' to determine the'nettron level'in the reactor..
The Startup, Instrumentation detector: signals are routed through Ortec Model 142PC preamplifiers, 'then Ortec:Model590A amplifiers. The amplifiers include single'channel analyzers to separate gamma pulses from neutron pulses. A Canberra Model 3125 Dual Voltage Power Supply provides power:toboth detectors. Processed signals are displayed Page 7

RAItResponse on individual Ortec Model 449-2 Log/Lin Ratemeters. The instrument,suite also includes an Ortec Model 994 Dual Counter/Timer,-.:
Two uncompensated ion chambers comprise linear power channels LPf ahd LP2. Each is powered from its own 300 volt battery. The detector signal is process*d by a Ciruit Equipment Corporation Model 1718 Linear picoammtieter.' These insruments have 9 ranges, from. 1 x 10 amps to6 1 x10 amps andidisplay current in amps. The.
picoammeters have an internal relay with a variable setpoint. The relay provides a high current scram signal to the rod scram circuit.
One uncompensated ion chamber comprises log power channel PP2.The ion chamber is also powered by a 300 volt:battery and, the signal is processed by a. Circuit Equipment:
Corporation Model 1718 log picoammeter. The instrument has, indication from 1 x 10q4 amps to 1 x 10 amps. Current in amps' a'd reactor perid in seconds~orin, decades per minute are displayed on the rmieter face. An initrmai telay sends trip sig*ials to'the rod scram circuit and has 'variable setpoints for high current atnd fast penriod. A second rIelay provides a control. rodoutmotion interlock at fast period: The period setpoints are 5.
seconds for.a scram and 15 second to blockcontrol rod; outmotion:
These instruments also have:reeorder outputgi;gnats, percent of full scale for LP1 and LP2 and amps! for PP2.       .       .
Four video graphic recorders are mounted, at-the main.control console.A Thermo Westronics Model SV-100 is used to display moderator temperature measured by J-type thermocouples. Two Thermo Westronics Model SV-180 recorders display LPI, LP2, PP2, SUA, or SUB, as the operator chobss. Typicaliy'onle would display the two startup channels on 6ne s&ceen aid the&#xfd;      o'ecodw6ild disp-lai,0P2, 6r'LP1 and LP2, depeiiding ipon the operiting range. A third recorder', 4I-neIyo        M ul~tiTrend Plus, is al so used and can display any of the same"cfi'niels. ;The-o;pat6r cafn have all three Stts of instrumentation iin ,View-on these three; re6o8d t&r.* Th8 &#xfd;reidrders als6ser'v' as data recorders by writing.to 3 1/2 irnch tfloppy, discs, thatare thehntransferred toa computer hard drive for analysis and storage. Data recording is independent of the screen view. All signals sent to the recorder are recorded. Temiperature on the SV-100 is recorded once per minute. Ion chamber currents, reactor peri6d and BF                  per secnd are recorded once per second.                                ..-   ,     .cc':t p*" seon are recorded One of the Thermo Westronircs recorders senses 1ow Stariup channel B. count rate and, implements the rod outmotion'interlock for this condition., The' setpoini is at ' counts per second.
7.3 Section 7.3, Reactor Control System. Provide a iore detailed discussion of instruments provided to'm'onitor Various reactor system processes and variables.
Examples include, c6ntf6lr6'd position indication, reactor temperature, reactor tank water level, reactor tank:water temperature, equipment status indication i(e.g",;air-Compressor) and various alarms, such as reactor tank leak alarms (

==Reference:==

NUREG 1537 Part 1, Format and Content Guide,: Section 7.3).
RPI Response - Instrumentation to. monitor reactor power level is described in. detail in the response to question 7.2 above.
Page 8

RAI Response Other indications availablein the Control Room are:
Control Rod Position, inches withdrawn as well as top and bottom lights Moderator temperature Reactdt TAnk water level Crea Monhinu (4) ridiationt levels Contiu'Asr Monitor count rate and volumetric air flow Rod magnetic' clutch currents:.,,
Dump valve solenoid current Startup source position Operating lights, for.equipment, specifically the air compressor, immersion heaters, reactor:fill ,ipmp, agitator motor, dtiinp'valVe position, fill valve, and drain' valve.
The operator has alarm lights on thelpicoammeters showing a scram.or a rod block. There is a rod block vitual alarm for low startup channel count rate on the videographic recorder that'impleffihts that interlock.
Control rod.position is. derived from optical encoders mechanically linked'to the rod drive gearing. The encoders transmit a series- of puls's to counters mounted on the control',
panel. The counters interpret the train of pulses to calculate and display rod .position to 0.01 inch resolution. Separate limit switches detect rod position at the topand.bottom limits of rod travel. The top limit switches stop outward rod travel and illuminate a top
"*~' limit light on the control panel/.Rod bottomil'imit switches activate Rod Bottom lights on the control panel.
Moderator temperature is measured at &#xfd;everal elevations in. the reactor tank by Type J thermocouples. T'drmocouple voltage is converted to Fahrenheit degrees by circuitry in the videographic    recorders that          tempprature. One  Oisp~ay recorder Screen displays three of the available themr'couples. A secnd        recrd,6er, displays ne thermocouple. N0,provision is made to measure reactor tem                      itc th     'as' mperait,   sin..c,js the, same'.,      yoderator'pe  Nopovso temperature.
Reactor Tankmwater levelis: displayed-inae&sect;ightglass 4in thefcontrol room.'
The four channels "ofthe Area Monitbring-sySte-tdisplay radiation levels in the Control
     .. Room, 'the EqUindnt Hall, O'tside~th~'~VueI' Vault, and on the ReactorTank upper deck..
All four'channels'have visual alerts and aiudible aarms.*
The Continuous Air Monitor samples air, above the Reactor Tank. Activity is displayed in the Contol 'Room and the instrument has an audible'alarm. Air flow is measured by a mechanical gage located in the ControlRoom.                -.
The power supply that provides direct current to the rod drive magnetic clutches displays current to each clutch. The same power supply provides power to the solenoid valve that regulates op&rating air to-the dumn p valve. Current to this solenoid is displayed ,at,the power s'uply. Power supply input andoutPut vdltages are also displayed.
All the operating ,lights are located on the Auxiliary Control Panel, CP-2, in the Control Room. Position of the startup source is displayed on CP-2. Other operating lights in the Control Room are on the Control Panel, CP-1, and indicate Shutdown activated, and Scram initiated. Shutdown is an operating mode that drives all rods inward at the normal operating slspeed. It is an interlock in that Shutdown mode overrides an outward motion command. The Scram light indicates when a scram has occurred.
Page 9

RAI Response 7.4 Section 7.3, Reactor&#xfd; Control System. Figure 7.1 shows four ion chamber inputs.
Section -7.4.states that there are three.' Clarifythe apparent discrepaincy and indicate the location ofthe detectors relative to thecore.
RPI Response - The:block, diagram isoutdated. 'Three ion 'chambers ar'e-in use and they ate described in responses toiiuestions 7.1 and 7;3. above. All three ion chambers are on the perimeter of the core at approximatelythe mi'dpland of the fuel. PP2,'connected to the log picoammeter islocated.between Rod 5 and .Rod 7. See Figure 4:1. LP1 and LP2 'are connected to linear picoammeters''ae located'. near Rod 4 ahdRod 3 respectively.
7.4 Section 7.6,, Control"Console and Display Instruments. Provide a' more detailed',
discussion of the instruments, controls, and indications provided on the main controlf console.
RPI Response - The main control panel 'is known as CP-1.*Electrical power for CP-1.
comes from the building lighting panel. CP- I displays include the position of the four control rods, including lights indication rods at the top. or rods at the, bottom. The shim switch' for the control rodsis, located on the.same section of panel as the rod position indicators. Each rod also has a switch.to select controlto te rod saimnscwitchnai    llor any one of the rods mamy be seiected.f Thus rod's' may be moved rnoany combination desired frnm thi- Qinah- Qhim qwit        ee thp nhntn. helnu, Page 10

RAI Response An adjacent section of CP-I contains the keylock switch to energize the panel, a keylock switch to energize. the scram circuit, a switch to selectshutdownmode, a switch to-:'
energize recorders, _and a &#xfd;scram'switch. -Indicating lights are associated with the shutdown switch and the scram switch. This section is shownat the right side.oflthe photo above&#xfd;:
A third section of CP-Lcontains selsyn dials for control~rodpositioh. This section is inactive since the 400 Hz.MG set that provided po.wer for the selsyn units was removed.
Above the operating icontrol are the electroni csfornthe neutron detectors. Abovethat row of instruments are three of the .videographic recorders.-A separate vertical section of CP-1 contains the Area, RadiationMonitoring displays and the fo.urth videographic recorder,,
The only operating controls on CP-1 are for control rod motion, including the manual scram. Control qf other pumps, heaters- and, valyesi are on, CP72 the ,auxiliary3 control, 7.6 Section 7.7, Radiation Monitoring System. An alarm setpoint for the CAM is specified; please relate.the setpointto the radiological impact:,.
RPI Response The purpose of the CAivM is to monitor for particulate activity in the reactor            asamarker   of possible*fuel ;element failure, rather than for radioogical 6%om, protectio'n purposes. The filter co'lectg fiss'ion products" which are detected by the 6M detector, so the precise radiolog'ical impact is inideterminate, depending upon the collection time and the degree.;of..eqvuibriihnmprfsent..;
XUp6h rreflection, it seemslikely &#xfd;-atthiss.svs crin, s .remnant
                                                            .a        of..the hUE core that Was eisent th;*le facilty pnioi-to 19, when fudl                               only 0 0054in of 7 clad byynent.were stainless steel:- Thd!4fi       of on,, :,'the'se elements resulting .in the ielease of contained fission products-is, assb rned to be `2,-"aorioeilausible occurrencebthan- the physical failure of oneeofnheSPRT finus                                      This equipment;&#xfd;wili beremoved from the SAR an "iany requirement for operation' removed fromUhe T&chnical Specifications.
,8.1 In gdneral'l this chtpterdoes not provide sufficient information to determine~the function and design basissof the Normal and Emergency Electrical Powerc Systems.
Provide &#xfd;a rnore:1etailecd discussion of the-desigiiL.sis and functiofial description of the' norinrl, and;emetgency:lectrical systems&#xfd;,Th & :,dnse should ensure that sufficient informn~ation lS&sect;provided .to address each of the applicable items listed in Sections 8.1 and.
8.2 of NUREG 1537, Part 1. Specific requests include:
: a. The ranges of electrical power requirements (1oftage, current, .. ouency4;
: b. From th0 ,verbal response during the sitV'is&#xfd;iit, it appears that a loss of normal AC power will res'ult in a loss of all lighting in the facility (withno emergency lighting provided), the'fire detection system, and the area radiation monitoring systems. If this is the ca&sect;&#xfd;e;.a justification should be provided in the SAR.to supporfthis design;
: c. How, instrurfientation'andt control circuits. are protected from electromagnetic inteffr&#xfd;cni     t ia* may be generated by the electrical power system; Page 11I

RAI Response RPI Response - The normal electrical power system consists of 60-heitz, 480-volt, three-phase power from the utility grid. The incoming service line is rated at 200 amps.
The 480 volt supply directly powers&#xfd; the fill-pump, .air compressor,.immersion ;heaters, building crane; agitator motor and a building air' conditioner. A 3,0 kva transformer reducesincoming voltage to 120 volts. and feeds' alighting panel. The lighting panel powers the rod drive motors,, source&#xfd; drive motor, lighting circuits andwall outlets, the two control panels;CP- 1 and CP-2, and the facilityfboiler'house. Instrumentation and recorders are powered from standard 120 volt outlets. The rod drive motors: and the source drive motor are the only three-phase loads on the lighting panel.
Facility po~ver usage is highest when preparing for operation due to operating the 2 horsepower fill pump to transfer watetinto theReactor Tank from the Storage Tank. The 2 horsepower air compressor cycles periodically during, startup preparations and during reactor operations. Unless the immersion heaters are in use, power consumption is very low during operation. The four rod drive motors are rated at 1/20 horsepower and few other loads are in use diring operation. When operated the immeision heaters draw 36 kilowatts and the 2 horsepower agitator is 6perated to keep the moderator temperature uniform. This is by far the largest load needed for reactor operation and is used infrequently. Immersion heaters are operated~for experiments that require a change of water temperatulre, f' examp~le, measurement of the moderator temperature coefficient of reactivity. Building capacity exceeds the demands.
All buildingmwiring was installed, to codes.. applicable at' the time of. installation. No known deviations are present. Repairs. to component failures are made withnew components. New motor starters;were instal'led.for the fill pump, agitator, air.compressor, immersion heaters, source drive motor and the rod drive motors in 2006. These are commercially available General Electric motorstarters. Themain! circuit breaker panel' and the lighting panel were both replaced in,2006 also. C.:.
No backup power is available'fOr the fa~iiit3. A         wyp  rvp Otage xiill imfiediaiely cause a 6u reactor scram due to loss of power to the rod drive electromagnetic clutches. Loss of sit6-power also deenergizes the solenoid that holds the dump valve shut. These actions place the reactor in the normal secur6 shutdovwrcondion - rods scraimmd and moderator drained from the reactor tank. Since the RCF does not generate appreciable fission products during operation there is no decay heat 'load to di'sipate. In a securesushutdown condition there are no'discharges ofi znyaterial from' th&#xfd;efaci.lit;y.
Emergency procedures for loss of site powerrequire the operator to-remove keys-from.,
the reactor control panel, CP- 1, and. t~uMnPff.instrumentation.to prevent power, surge ''..
damage if power recovery is erratic. These actions will maintain the secure shutdown condition when          is restored.
Emergency lighting for safe egress is installed in'the appropriate areas (installation completedJune 2008). Battery powered' flashlights are'also 'available..
Loss of power also disables all monitoring systems at the facility such as fire detection and building security. In this situation standard procedure requires continuous, on-,site surveillance, normally provided by RPI Public Safety.                  '
Instrumentation cabling is shielded to protect from electromagnetic interference.
Page 12

RAI Response
: 9. AUXILIARY SYSTEMS 9.1 Section 9.2, Handling and Storage ofReactor Fuel. This section references a constraint from -the.design basis for the fuel vault 'which places a. limit, of 15 fuel pins per tube in the vault. Section 1.3 states that; the'vault, has short tubes, forithe former fuel design andlong tubes-for the currentfuel design. Is'it possible to place the current fuel in the short tubes or more than 15 fuel pins in a tube? If either or both are possiblei please' discuss the consequencesof such 4n accident.,
RPI Response- Both are possible.i The'cbnsequences of placing Apin in a short tube are trivial. The pins. will simply extend beyond: the end of the tube. Otherwise, the tubes are structurally identical tothe, long.tubes.       ..
Consequences of overloading the tubes (more than the limit of 15 per tube) are non-trivial only in'the event of a massive flood'in which the entire vault is inundated. Recent Monte Carlo analysis u'sing'MCNP [Ref] assum~d that that the vault was aninfinite arraf' of fuel storage tubes, completely flooded with water.The' ahalysis showed that the maximum reactivity in the arrafigeiment was reached att 53'iins.       eThemaximum infinite         ,
multiplication factor, kif, was less than 0.6900             'ompared to the 15 pin case "where kin'f' 0.6100.
9.2 Section 9.2, Handling, and Stbrage of Re actor!Fuel. SAR Section 9.2 and TS Section 5.6 describe the storage of spent fuel and: the surveillance requirements and frequency for.
fuel inventory., Is this surveillance a TS requiirnement;and if not, justify why it is not?
RPI Response Inventory requirements, are derived from the government ownership of the fuel and our iequirement to report inveht'1ry annually to the Nuclear Materials Management and Safeguards System.(NMMSS) &#xfd;It is not a. necessary Technical Specification requirenient.,'".            "       "   -
: 10. EXPERIMENTAL iAACIL1TiES AND UTILIZATION, 10.1 You may not have experimenta! facilities such as~those listed in NUREG 1537 (Part 1), Se&tidn 102, however, the second p~aagr.aph on page 10-2, concerning critical, facilities, is applicable. In addition, the paragraph just before Section 10. 1 and other parts of this chaptei of NUREG1537, (Part: 1), incbudirigthe Apperndkies 10.1 and 10.2, concerning experimental utilization are alsouapplicabe. ,             -
From the list of experiments described in Section 1.6 of the 'SAR it!x'ould appear that experiments. performed.at the RCF are limited to the measurement of reactor characteristics (rod position measurements, subcritical, multiplication .measurements, etc.). However, during the site visit reactor use was characterized as "...used for demonstrations about 95% of the time. Periodically goidIfoiis are &#xfd;activated, butt only to a level that d'oes not require the' foils to be plaCed in lead pigs for transport. (Core flux is not high enough to highly activate the foils). All experiments' are reviewed by NSRB.'
Page 1-3

RAI Response Discuss your experimental program,, including information which more:fully describes.
.the types of experiments performed and the tfacilities, apparatus or equipment used to perform them. In addition, describe the process for experiment approval:and oversight.
Please use the references mentioned above for guidance.
RH Response -"Currently, the primary .                         of the reactor i- to suppoittthe Critical Reactor Laboratory class and p&#xfd;erforfnrderrinistrafonKs. Dbenmonstrations cronsist of one of the approved experimefits derni'nstrated for smnall grotip of,'studenis who are visiting the laboratory as part of a familiarization touri These are RPI studentf'*'ob may take the formal laboratory course in a subsequent semester. Demonstrations may also be arranged
                          . .I ... universities, forea students"". from. other      . '.* ."n   r although
                                               '"a o ' s,'e' ,,'ira  "Ai " &#xfd; ', not":,occurred
                                                                .' that.has             al'- *        ."e' past in the    i . ' several t e w r years. The difference between a demonstratin and the formal lab c6urseis iethe work expected of the students. The lab course requires formal reports and more active student participation in dat' recording and analysis.
The typical menu of experiments is provided below;.:
Source range channel calibration,(BF3 detectors),.
Fuel Pin, Addition -Approach to Critical using Inyerse Subcritical Multiplication Plots,,,                                         .   .
Exact Bank Critical Rod Position and Excess Reactivity Measurement Bank *nd Individual' 'Control Rod" "WrthM1asurment, measurement of IndividudiaFuel Pi"W0rt-s, "
Isothermal Moderator TempeeiatureCoefficient of Reactivity, Void Coefficient of Reactivity,                           "'"
           ... BCoeffic.ient.of Reactivity,                  '.
Interior/Exterior'Radiation Survey atiPower,..,;.,,..                               .              ;
Axial and Radial PowerMapping,: ..- ,                             .
Power Calibration using Gold Foils.
These are pre-approved experiments and are documented in the MANE-4440 Laboratory Manual.: No equipment or facilities outside ;of the RCF-are, required for these;-. :.. "
experiments. The RCF has gamma-spectroscopy~equipment&#xfd; used tO analyze the gold.
foils and perform gamma-ray scanning of pins, boron-impregnated.tape and polystyrene:,
material are used for the boron and void coefficient studies, and electric immersion heaters, are used for the, moderatorkdensity coefficient: .
addition&#xfd;l&#xfd;ekperiinents' htav{, b:e'e cInd          cted
                                                        ..          e.g., TLD .chip actvations; ahd! otlhets a-re planned. Inmall'cases, the ope'ating'procdtres' are folldx6d depending on wh&#xfd;e'ther the' new configtiratifn is'a "knOwn":o-or "unO1W'xn" :c6re.
Additional tests planned include criticality benchmark testing of partially-reflected core configurations (using Zircaloy, silicon carbide,,concrete, aluminum, steel, etc), and.
Borobond TTm ..                                                                                                 ,
All new proposed experiments and procedures are reviewed by the NSRB, in accordance with Section 6.4 of the TS.                                                     '
Page .14

RAI Response 10.2 Section 10.states that new experiments that:raise aUSQ will be reviewed by the:
NSRB'. Note that under.10UCFR Part 50.59 this will require a license amendment. Please resubmit wording for the referenced -paragraph-incorporating the. current wording of 10CFR Part 50.59.                    ..        .
RPI Response - RPI will modify Section 10 of the SAR toread, "All new experiments or classes of experiments that raise an unreviewved. safety question' shall be reviewed and approved by the Nuclear SafetyReview Boaid in.accordance with 'Section 6.3 of the Technical Specifications. 10 C'FR 5.59 Will be.o    Clsulted to determine is a license amendmeht is required."
: 11. RADIATION PRQTECTION-PROGRAM AND WASTE MANAGEMENT' 11.1 Section 11.1.5, Radiation Exposure and Dosimetry., Discuss typicl .dose rates throughout the RCF during reactor operation, fuel handling operations and shutdown so as to give a perspective of the radiation environment,..
RPI Response - Dose rates at the RCF during reactor operations at full power are discussed in iresponse to SAR RAI 4.,8. This is an extreme case-foi the facility; since typical operating power levels are far lower than the full license powei-.1 After reactor shutdown, the maximum dose' rate in the facility is at the position on the deck above the reactor tankwh ere dose. ratqs range fr6mn about 5 mR/hr shortly after shutdown to about 0.3 mR/hour we~lipast shutdown. Do .se rates quickly drop to near background a few meters from the reactor tank.         .
Specific surveys have not been conducted during fuel handling procedures. HOWever, the quarterly accumulated dose measurement during a' recent quarter where a full core unload and reload was performed revealed that no staff member exceeded the-nminimum detectable dose of 10 mrem for the quarter.       .        . .      .  ,
11.2 Section. 11.1.5, Radiation Expogure. and Dosimetry. 'Please provide; information,,
indicating thatthe radiationo levels at the site bodndary~are within the regulatory'limits during andafterreactor operation.:
RPI Response - Environmental monitors latthe site boundary are subject to a minimum detectable quarterly dose of 10 mremr, and have rarely been shownto exceed this. value.
Environmental monitoring. results ae reportedinithe. annual operating report, and are consistenti yshown to be well below                      .egulatoryliims.
See also theresponse to SAR RAI 4.8..
11.3 Section 11. 1.7, Environmental Monitoring.: It is"stated that 5 mreimn/yt has been measured at site boundary and 15 mremlyr at the exclusion area boundary above' that; measured at the GE facility more than 1.6 kminaway. The staff is reading thisas:
5 mrem/yr above background at the site boundary and 15 mrem/yr above background at the exclusion area boundary with the background taken at the General Electric Company Guard Station. During the site visit it was stated that this was old information and that recent results reported in the annual effluent report for the RCF indicate no detectible Page 15

RAI Response radiation at either the site boundary. or the'exclusionr area boundary. First;, if there is more recent anid accurate environmental monitoring data available, please. provide,.an update for section 1 1.-7 of your SAR; iotherwise discuss-how you satisfy the requirements of 10 CFR-20,1:1.01 (d):and verify that you meetthe requirements of' 10: CFR.20.1301 (a)(2).
Second, clarify the discrepancy between the above statements and TS 5.1:and TS 5.2 which indicate that-the exclusion areaiboundary and-the site boundary, are. both defined by the outer fence surrounding the reactor building.
RPI Response - The origin of the environmental monitoring, values stated is unclear, as those values are below the minimum detectable dose for current or previously available environmental, monitoring devices: used at-the facility. The current-environmental monitoring data is, reported inSARRAI. 1.12.-Additional discussionis.-provided with the response to SAR RAI 4.8.        -,     .                      ,   -
Compliance with 10 CFT 20.110.1(d) is~demonstrated:through annual reView.using the "
COMPLY code and conservative assumptions regarding Ar-41 generation in the target room.. Compliance with 10 CFR 20. 1301(a)(2) is demonstrated by the values reported in response.to SAR RAI 4.8            .           *    -                          ,   .
The references, in TS 5.1 and 5-.2,are correct,, and the SAR will be updated to reflect them.
12.: CONDUCT OF OPERATIONS .(INCLUDES TS.SECTION 6, ADMINISTRATIVE CONTROLS). -
10 CFR 50.36 contains the regulations for technical specifications. 10 CFR 50.36(b) states that the TS will be derived from the analysis- and, evaluation included in. the safety .x analysis report. However, SAR Section 12 is quite brief and in many sections just refers to the TS, which is reverse from-theintent-ofithe: rergulatioins. Please resUbmit Section 12 of your SAR, addressing; each: of the-issues identified and-questions raised in the following RAIs.       .                    -      ,   -    .                .
RPI Response    -,Chapter. 12  has been rewritten.       .
12.1 SAR Section 12.1, Organization, and TS Section 6.1, Organization.. NUREG-1537 and ANSI/ANS-15;1-1990, "The Development of Technical Specifications for Research Reactors," provide, guidance-on the Organizational structure, The guidance notes.that there should be a multi-level organization chart in the SAR and a description of the relationships with the line. organization. The SAR andTS contain such descriptions and charts, however, the charts and titles -are not-completely consistent, e.g., itappears that    -
the operations supervisor, reactor supervisor, and supervisor-of critical facility and:,     -
radiation safety officer may be one and-the same. 'Please. clarify and make terms agree between the SAR and TS descriptions, and the Figures..          ,             ....
RPI Response -See the revised Chapter 12 and Technical Specifications Page 16

RAI Response 12.2 SAR Section 12.1, Organization, and TS Section 6.1, Organization.: ANSI/ANS-15.1-1990, defines the responsibilities of the.Level 1 Management position as responsible for the reactor facility's licenses or-charters(i.e.,-Uniit or:Organizational Head). Verify that the RCP Directoi has authority and responsibility:and speaks for :RPI in all matters concerning Licerise CX-22. As an example,, decommissioning funding is requirediby 10 CFR 50.75 (e)(1)*and tpically the Level 1,'Manager has authority to provide the financial assurance required by the regulations....           .......
RPI'Response      See the revised Chapter 12      .
12.3 SAR Section 12.1, Organization, and TS Section 6.1, Organization. The organization; illustrated' in Figure 12.1 of the SARis ,different than that in Figure:A1.of the TS. Please resolve those differences and justify the structure.
RPI Response - See the revised Chapter 12 and:Technical Specifications 12.4 SAR Section 12:1; Organization; and TS secti6n 6. 1, Organization. TS; 6.2 states that the Nuclear Safety Review Board (NSRB) advises the Facility-Director, TS 6.2.2 (a) states that the Chairman of the NSRB is approved by the Facility Director, and the SAR Section 12.2 has NSRB auditreprtts going to theFacility Director,.whereas SAR Figure 12.1 and TS Figure 6.1 show the NSRB reporting to the Operations Supervisor. Clarify the relationship between the Facility Director and the NSRI such'that independencelof.
the review and audit function of the NSRB is assured. The ANSI/ANS-15.1- 1'990 and NUREG-1537 provide guidance that may be helpful.
RPI Response- Seethe revised.Chapter.12: -                ',
12.5 Section' 12.1:3, Staffing,! and TS-Secti1n -6At.3,;Staffing.&#xfd; "ANSI/ANS'4-15.1-1990 :
provides definitions ;of reactor. se'cured amnd~reactot shutdown. TherTS provide' similar, definitions for "reactor shutdown," (TS 1.3.0), and "secured shutdown" (TS 1.3.U).,,' The TS only specify the minimum staffing when the reactor is not shut down. Thus, the TS do not specify the required staffing when the reactor is'shut down,4but not secured: shut down. Propose a TS that specifies the minimum staff required when the reactor is shut down, but not secured shutdown. "          ..    !.          .      .' .  '           5 RPI Response - See the, revised Chapter 12'and Technical Specifications 12.6 Section:: 12.1.3, Staffing, andTS, Section 6.4.3, Staffing.- ANS-15.1-1 990" recommends 'forthe SRO to be capable of getting to:the r'eactor facility within a reasonable time (e~g., 30'minutes). The proposed.TS 1.3;P defines "Readily*Avdilable on Call," used in TS.6.1.3(a) (3) aswithin30rmiles :or6 minutes;. The existing TS 1.3.P defines, "Readily Available on Call," 'as 1.5 miles or 30 minutes. Please justify 'that 60 minutes is an acceptable response time for the SRO readily available on call.
RPI Response - The ANS Standard gives 30 minutes as an example, not even a recommendation. For RPI, the 30 minute or 15 miles has been an unnecessary burden at Page.. 17

RAI Response times, and.no such prompt response is judged to.be necessary&#xfd; Phone contact is adequate' for immediate assistance while2 a second operator arrival up to. an hour later is judged adequate response time. The available personnel can place the-reactor'in secure shutdown in about one minute and wait for arrival of another senior reactor operator. Emergency assistance for fire, injury or security issues is.a few. minutes away and provided by Schenectady civil:authorities or local ambulance services. RPIzCamrpus, response such as Public Safety or radiological assistance is about 45 minutes away. The requirement is changed.to 60 minutes and 25 miles. Seethe, revised Chapter 12 and Technical-Specifications.
12.7 Section 12.1.4,` Selectidn and Training of Personnel, and TS Section 6.1.4, Selection and Training of Personnel. The TS cites ANSI/ANS 15.4-1977rather t-han the more recent version, 1988. Please update this reference ifposs'ible, otherwise discuss the' reason for not updating it.
RPI Response - The ,reference has been updated.
.12.8 Section 12.1.4, Selection and Training of Personnel, and TS Section 6.1.4, Selection and Training of Personnel. Discuss. how your training program meetsthe requirements of 10 CFR Part 19.
RPI Response. - All staff receive training as participants in the RPI Radiation Safety Program, which includes 'initial'training ahd refresher training at least annually. The training program is in accordance with the New :York State Department of Health regulations (State. Sanitary Code Part 16), since that-is' a'uttirity which licenses RPI's use of radioactive materials, and which are at least as restrictive as the 10 CFR provisions.
12.9; Section- 12.1.5, Radiatidn: Safety. i0 CFR 0.1101 requires that each licensee shall' develop,:document, and implenment a radiation protectloniprogram. The NRC staff must have adequate information about your radiation protection program to be reasonably assured that it meets the requirements of 10 CFR 20. NUREG-1537 and Section 6.3 of ANS-15.1-1990 recommend a TS on Radiation Safety and ANSIIANS-15.11-.1993, "RAdiation Protection' at Research Re:actor Facilitie," provides guidance.. Currently the brief descriptionfs of the radiation safety organizetithi in the. sAR and TS are not coordin'at:d arid do not usse'the ,same term's.' The Radiation Saft 'ft rms.'he~aiatin Saetyfunction is not included on Figure A. 1 of the TS.: Ini' Figure 12. 1-of the SA&sect; "'the Direcitr, Office of RaIdiation and Nuclear Safetytis connected to 5 levels of theorganizaton 'without any'.description of chain of command,' reOpotti'ng, 'coordinati'n, etc. The health phys'icist' of the TS is not mentioned in thle SAR', soit is. not'cledf where the person resides within the organization.
There is no commitmient to ANSIVANS-i15. for* mentionof an, ALARA program &#xfd;in the TS. In addition, there is no discussion of how and when the radiation safety staff.
communicates with the facility manager and Level 1 management to resolve safety .
issues.
10 CFR 20.1101(b) requires an ALARA program. Who is responsible for the.ALARA and radiation safety programs?? When the RCF is in use, is there-a perso n responsible for radiation safety present at' the facility or on call? If on call, does the SRO have sufficient Page 18

RAI Response training in-radiation safety to perform thosbeduties until assistance arrives? SAR Section 12.1.5 suggests thatRCF staff have responsibility for radiation safety and: the campus support is onlyavailable foroccasional assistance.           -
Please add discussion in the SAR to address the above questions -and issues. As appropriate, propose TSs'and supporting bases1that reference. discussion in the SAR.
RPI Response -The radiation safety program for the reactoiris under the pur'iew of the RPI Office of Radiation and Nuclear Safety (ORNS), part of the Environment, Health and Safety Department, as part of a radioactive materials license issued by the New York State Departmentof 'Health.,            -
The revised Chapter' 12 and Technical Specifications have corrected the organizational chart and harmonized the terminology related to the radiation safety)program.'
The ALARA program is part of the campus radiation safety program, and is described 'in the campus Radiation Safety Manual. The Radiation and Nuclear Safety Committee is responsible for reviewing the programanid ensuring thafiadilog'ical operations on campus, including 'at the reactor facility, remain ALARA.
During regular operation, the SRO is responsible for normal radiation safety tasks, such as the examplesprovided in 12.1.5. A dedicated member, of ORNS, need notbe present at the facility. for.these, types ,of measurements. ORNS&is. always on call to respond to emergency situations, and theRadiation, Safety, Officer provides oversight, assistance, and support for the radiological aspects of facility ,operations.
A new section 12.3.3 has been added to, the SAR discussing the inclusion of radiation protection procedures in the camp s.Radiation Safety Manual. See the revised SAR Chapter 12.,                ,            .        ,
12.10 Section 12.2,'Review.and Audit Activities; and TS Section 6.2.1, Composition and, Qualification. ANS- 15.1-1990 states that members and alternates of the review/audit cmite s'houdb committ e' hld'be appoIne            by an'd appointed by       r and report-to Level 1I management, the level above the individualresponsible for facility operation. NUREG-1537 states that members should be appoint'ed by the' highest level of upper management., However, this does not appear to be thee' casein the SAR, the TS, and the Organizationai Charts,]Figures 12.1 of the SAR and A. I of the T8. Please di cu s's and provide assurance that the NSRB is independent of the direct' management of the facility. Propose TSs a*id SAR bases as necessary to address this'issue (see also 12.2 above).
RPI Response      -S see the revised Chapter 12' 12.11 Section 12.2, Review and Audit.Activities, and TS 6.2.3, Review and Approval Function. NUREG- 1537 suggests that the TS should explicitly state that the NSRB,.
addresses' thi review functibn of 10 CFR 50.59. Neither the SAR nor TS explicitly mentions this function. During the site visit, the licensee noted that the NSRB does Page 19'

RAI Response perform-this review function. Discuss how-the 10 CFR 5.0.59 process is implemented at, RCF..
RPI Response - This question is addressed in TS 6.4.1. This paragraph has been, amended to state that the review function of the NSRB is pursuant to 10 CFR 50.59.
12.12 Section 12.2, Review and Audit'Activities, and.TS 6.2.4, Audit Function.
NUREG-1537 and ANSIIANS- 15.1-1990 note areas that should be addressed by the Audit Function. Items listed~there, but, not:explicitly included in the SAR or TS audit function, are: TS conformance, thephysical security plan, requalification training ,
program, emergency plan, and radiation protection program. ,Provide assurance that these areas are part of the auditfunction of theNSRB.             .   .
RPI Response: - The, TS has been amended. to ensure that these areasare addressed as part of the NSRB audit function;.               ,..
12.13 Section 12.3, Procedures and TS Section 6.3, Procedures. ANSI/ANS-15.1-1990 lists the activities that should,,be addressed by written, procedures,. Provide justification for not including personnel radiation protection procedures, including ALARA during normal operations per ANSI/ANS 15.11-1993, and administrative controls for operations and experiments il TS 6.3..
RPI Response - Since the RC1Fhas adopted, the campus radiation safety program as its radiation safety program, theradiation.Pmrotectionrprocedures are administered under the structure establishedby the radioactiyematerials license with the New .York State.
Department of Health., As, such,,,the procedures are:inco rporated by reference and not included specifically in Section 412.3. However, we agreethat "Radiation Protection should appear in the bulleted list in section 12.3:,$eetherevised Chapter 12..
12.14 Section 12.3, Procedures, and TS Section 6.3, Procedures. 10 CFR 50.36(c)(5) requires administratiyveTS. :NUREG-537 and!ANSIIANS- 15.1-1990,-Section 6.4, provides guidance for meeting the requirements for review and approval of procedures.
The SAR and TS do not discuss -how,facility operations and management prepare, .-.
review, and approve the, procedures: Discuss yourreview, and approvalprocess to.
provide assurance that there is adequate independence,.                .   .
RPI Response        See the-revised Chapter J12 12.15 TS Section, 6.4, Experiment Review ,and Approval., Reword this TS'utilizing the terminology *of the present version.of 10! CFR 5Q.59,.
RPI Response - See the revised Technical Specification.
12.16 TS Section 6.4, Experiment Review and Approval. RegulatoryGuide 2I2, "Development of Technical Specifications for experiments in Research Reactors," 1973, Page 20

RAI Response provides, guidance in meeting the requirements'of 10,CFR 50.34(b)(4) and 10 CFR           .
50.36(b) with respect to theexperimental program. Provide adequate discussion and propose TSs as necessary to allow the NRC staff to assess the risk to the health and safety of the public from:the: operation of your facility.
RPI Response - The proposed changes to TS 6.4 and the description of the experimental program are provided in response to SAR RAIs i0.1, 10.2, & 1.2.15.
12.17 ,Section 124, Required Actions, and TS'6:5, Required;Actions. The regulation 10 CFR 50.36(c)(5) requires administrative TS. ANSI/ANS-15:1-1990 provides guidance*.
and the RPI TS*define actions, tol be taken in caseof a reportable occurrence, the actions include "reactor conditions shall be returnbd to normal or the reactor shall be shut down."
NUREG-1537 Chapter 14, App. 14.1, Section 6.6.2 states that the TS should establish in advance specific Criteria.for the two alternative actions; return to normal and shutdown (an example is given. in the reference). Discuss the criteria used at RCF and'propose TS changes as necessary.
RPI Response The TS has been amended'to require thatfthe reactor be shut-down in the event of any reportable occurrence.
12.18 Section i2.5,'Reports, and TS 6.6.1, Operating Reports., The applicable regulations include 10 CFR50.36(c)(5)&(7). ANSI/ANS-15.1-1990 Section 6.7.1 suggests a list of those items. for inclusion in the annual operating report. The TS includes these with the exceptionof major preVentive maintenance and. a summary of.
exposures over.25% of allowable for-visitors : ANSI/ANS-151-1990 also calls for a..
summary of environmental surveys perfornmed outside the facility, -butthe TS only lists TLD dose rate readings. Arethere other environrhenta-tresultsithat should~be included?
Also TS 6.6r1(a)(5):and (e)Kcorrdctly),cfted 10 CFR'50.59,.but "(ad" and "(b)"
respectively, should be dropped from the 10 CFR 50.59 citations.
RPI Responsehs-See the revised'Chaptert,2,and Technidcal Specifications' 12.19 Sebtion 12,;5, Reports, and TS'6.6.2, Non-Routine. Reports., ANSIIANS- 15.f-1990, Section.6.7.2, specifies a 30-day report for permanent changes in the Level 1 or 2 facility organization, but the TS include thisas anfannual report. 10 CFR 50.36(c)(7) and the guidance in NUREG-1537 Chapter 14, App. 14.1, Section 6.7.2, Special Reports, states that the telephone reports should be made to the NRCI Operations Center and the regional staff. Written reports fall under 10 CFR 50.36(c)(5) and should be submitted as specified in'l.0 CFR 50.4. Propbse TS: to *equire a 3O-day report notifying the NRC of permanent changes in the Level 1 or 2 facility organization,., Propose changes to the TS so that all written reports are submitted as specified in the first paragraph of TS 6.6.
RPI Response - The Technical Specifications already state that all written reports shall be submitted accdrding to the first paragraph of TS 6.6.
Page 21

RAI Response 12.20 Section 12.6, Records,, and TS 6.7, Operatifig Records; The applicable-regulation' includes 1.0 CFR-50.36(c)(5):* The following! records specified in ANSIANS-15.1'1990 should be added .to the TS-listing" fuel receipts (5 years),. approved: changes tO -operating procedures.(5 years);&#xfd;,NSRB.;audit reports (5.years);. training'records Off: ertified operations personnel (one certification cycle), radiation exposure forvisitors (life of' facility).
RPI Response          -   See the revised Chapter 12 and Technical Specifications.
: 13. ACCIDENT ANALYSIS 13.1 Section 13:1.5, Mishandling or Malfuiti'on 'ofzFu1.el Section 4.5 states that "...
removing multiple fuel'pins from the interior sections 6f the core can result ini significant reactivity ddltioh, beyond the excess reactivit&#xfd;,limit oft                                of 60'cents setinheTcn nt the.Technical, Specifications. Please'provid ejustification ini Section                                     13.1.5  to support,the        statement that mechanical rearrangement of the fuel'to 6btain a supercritical configuration inadvertently or with intent, is not a credible occurrence.
RI Responise,- The procedures governing changes to known cores and the training of the SRO s prevents the irindyertent arrangement of fuel to achieve a suprcritical configuration' from being a credible occulrTence. The staffin reqirements specify at least two operators, "oneof which mUStlbe an SRO, be physically at the RCF in order to operate. 'The possibility. f bbth operators agreeing to rembov fuel to achieve a supercritical configuration is no' judged to' be credible.
13.2' Section 112, Accident Ana lysis and Determination of Consequences. In Table 13.1f;the ratios b/bl'ff 'appear .t6 have come frm G..                                          eepin, Physics o N'uclear Kinetics,"' 1965, butithe *value'o-f beff ='0.00765 isdiff&rent pfrin the y                                             of mucear
                                                                  ; ' "*'  '      ." [-. 'I"  `,:"
                                                                                                "'i[    .j 5.._   -. 'i  *:        .i I: .'.',
             * " " 1*",       , ',.,d            ,  ' ,,;q:
book. Please expiain'ho'w beff was determinedId...
RPI Response - The delayed. neutron fraction, 03, is a physical property of fissionable matenal! Several resourcesprovide valuesof 136r various fissibnable mateials,.the.
most notable being Keepiii's Work (G. R.' Keepih, "Physics                                           Nu'lear iie-Ph'&si" 1965)."
Th vialu'eof P3eff; however, is dependent 61ith e neutron .spectrum and fuel systemr" (fissionable material and enrichments) of thereactor. Itis essentially the 'spectrum-adjoint weighted3 effect of delayed neutrons, j:          fU dr Q' dEJ Xd* (E)'Nj(r,a ,E')vyd(r, E )E*(r, E)*(r, E dE- , V
:.   ,, (1).,
Ie J.
lf( _
fj J2 f   d 33 .f j jdr ij f ~,JdE.Jf d   t;, .    .    ..
x(E')f(r;,-E')v(('t,
                                                                "         ) v',' .. E)f (r,. E)O(r, r    . .E)dE .  . ..
where the numerator is the spectrum-adjoint weighted neutron production from delayed neutrons only and the denominator represents the spectrum-adjoint. weighted neutron productiohp from prompt and delayed neutronss.
This equation can be solved using deterministic transport, provided the delayed,&#xfd;Land v are' known. These values are generally available from ENDF/B formatted data sets.
Page 22

RAI Response Initial- diffusion theory analysis of the LEU core resulted in the current P3eff = 0.00765.
As a check, anew analysis7 was performed using. MCNP5"and the ENDF/B-Vl,8 data set:
Two runs are necessary. to calculate P3eff. &#xfd;The first irun is a normal iterated source eigenvalue calculation. The second iterated-souirce eigenvalue calculation turns off delayed neutrons.. [3eff is then given as keff ota1      keff prompt                               )
keff total where since continuous energy Monte Carlo is used, the adjoint weighting in (1) is approximated with the next-fission probability function. TheMCNP runs resulted in a p3eff 0.00813 +/- 0.00023! It is knwvn thatit'h ENDF/i3-VI_8 delayed neutron fraction for thermal fissi6n min U is 0.0069, or approximately 6% higher than Keepin's value..
Reducing the calculated value by 6%'tonmatch&#xfd; Kee-pin, the new calculated P3eff=0.00766,,
which is in excellent agreement with the current specified value.,
13.3 Section 13.2, Accident Analysis' andDeterniination of Consequences. Section 13.2 is the SAR info..iation which supports TS 3A1 and 3.2. While the TS appear reasonable, they are not fUlly supported by, 6r.consistent With ihe SAR 'the analysis is done at 20T C while'the'TS minimum allowable ternperatureis 5-T. F (GOT C) where there is a greater positive reactivty coefficient. Is,the2T C,calculation a worst case for other reasons? Since there is'no inierlock on temp rature, please. discuss why the initial temperature used in the'ahalysis hfiuldnOt bees,. han theS-,T FTS limit.
RPI Response - The analysis of Section, 3.2,was performed at,20 C because that is the temperature that the cross section fibray*       ,waspi.rided,      It is not a worst-case for other reasons. 1Ioweverthe.overl effect on' the*postulated accident scenario would be negligible if initiaed at 10 C'instead 'of20 C, as documented jn Amendment 7 to the operating license (July 7, 1987), Section 3.2.
13.4 Section 13.2, accident Analysis and Detennation of Consequences.. Section 13.2 begins by.stating the i-eict6r was 0perating at 200 watts at the start of-the scenario. Later it says thatTable' 13.1 lists nuclea characteristics used in the analysis, but is inconsistent in that it states the power to be 100,watts. Please clarify.                                         ,
RPI Response - We assume that the indicated power is ,100 watts. To, increase the conservatism of the'scenario, w'e then'aS'sumne that the indicated power could&#xfd; be off by as much as 100% from the true power. Table 13j11 wil~lbe updated'speoifying an initial power of 200 'atts.
13.5 S ection '13.2;,Accident'Analy is and Deternination of Consequences. SAR Table 13.2 lists a column of TS values which in some ciases differ from those used in the TS.
For example, a Shutdown Margin of >0.02 (2.6 $) is stated in Table 13.2; the limit stated.
in TS 3.2.2 is 0.7 $. The limiting reactivity 'iworthof a standard fuel assembly is listed to be <0.039-(5 $);t TS'3.2.1 specifies a maximum of 0.20 $. In the first example, the TS value is less conservative than the value listed in the table and therefore, appears not to Page 23

RAI Response meet the basis in the SAR however, the problem appears to be with theterminology. The value listed in the table as Shutdown Margin appears to be Shutdown Reactivity as defined in TS 1.3 V. The limit in TS 3.2.2 is consistent with the table value of "Reactivity with One Stuck Rod," however, the table does not specifyithat the 'stuck rod is the most reactive rod. In addition, the accepted definition for "Shutdown Margin" (see the definition in Section 1.3 of ANSI/ANS-15.1-1990) is not specified inthe.TS. In the second example the TS is more~conservative but the disparjty is so large the basis is questionable. Please provide more discussion about Table,13.2 including the source of the values and their relationship to the TS. Clarify th&.confusion with "Shutdown Margin" and "Shutdown Reactivity" by providing a definition of the former in the TSs and correcting Table 13.2to be consisteit with the'definitio'hs:, i(Se&*TS RAI 1.3.V)
RPI Response - A definition of shutdown margin has been added to the TS.
Table 13.2 will be updated with the following correctfons.
Shutdown Margin - remove this value from the table. The intent is: to specify.shutdown reactivity & shutdown reactivity with most reactive rod stuck.
Shutdown Reactivity                       ,           <-$1.00 (LEUValue) <-$1.00 (TS)
Shutdown Reactivity with most reactive rod stuck < -$0.65 (LEU value) < -$0.70 (TS.
Reactivity Worth of standard fuel- assembly.          <$0.039 (LEU value) < $0.20 (TS.
Other changes to documents:;
SAR, 1.3                                         -
"...Excess reactivity with al control rods fully withdrawn is typically less than 30 cents.
The minimum shutdown reactivity of the'reactor is a dollar., A more detailed description of the reactor is given in Chapter 4."
SAR 13.2
"...conservatively assumes the instantaneous insertiodn, of $1..000 negative reactivity (the, minimum core shutdown reactivity) at 5 seconds after the excursion begins."
13.6 Section 13.0, Accident Analysis, Figures 13.2 and 13.3. Notes on Figures 13.2 and 13.3 infer that the analysis was done for a 421-424 pin core with an 0.585 inch pitch whereas SAR Chapter 4 describes a' 3294to 333 pin core with a: 0.64inich pitch eore lattice plate. However, Chapter 4 states that other approved lattice-plates exist. In the SAR: and possibly the TS clarify what constitutes an "approved" lattice, the approval process and why the safety analysis presented e' i6pes"other lattices.
RPI Response - There are other lattices plates that have been installed and tested in accordance with the TS requirements for new experiments, with the appropriate NSRB approvals. The approval is contingent on an analysis indicating that the current SAR remains valid for the proposed lattice. Previously used lattices include the 0.585" pitch lattice plate referred to in the SAR. The most commonly used lattice plate at this time is the 0.640" pitch lattice. The references to "other approved lattices" will be changed to reflect the revised TS 6.4:
Page 24

RAI Response SAR 1.3 changed' to read
 "...in an octagonal array with a 0.64" pitch. (other. configuiations approved in accordance with section 6.4of the technical specificatioins exist as well) with 4 boron flux-trap control rods..."
SAR 4.1 changed to'read,
 "...The moat coriimonly used fuel pin configurati6tun'tilizes a'0.640" pitch (other lattice plates approved 'in accordance with Section 6.4' of the technical specification are available)cdntganing 329-333&#xfd;uel pins... .:                                        .
The proposed lattice-plateis analyzed to ensure that the current SAR analyses are valid*
for the new configuration. This is part of the approval process that is specified in TS 6.4.
.13.7 Section 13.0, Accident Analysis'. 'Pleasg*address the following editorial observations:
a.' Near the end of Table 13.3 it siates thdt the .temperature coefficient is negative when T< 16 or T< 32 for core A and B respectivelfy. This does not appear to be correct or consistent with Table 1312'and Figures 13.2 and 13.3.
RPI Response - The end of table 1:3.3 will be changed: in: the next SAR update to:,
Isothermal Temperature Coefficient for LEU Core A:.
aT(0 C) = 1.825x10-T 2 - 4.8x10-6T + 6.932x10-,
and aT < 0 for T > 160 C (61&deg;F)
IsothermalTemperature Coefficient for.LEU Core B:
c.T(0 C) =,2.A13xl07T 2 -5.0x.10-6      1.423x 1.04 0
and aT <'O for T > 32 C (91 F)0               '. -
: b. On Figures 13.2 and 13.3 the temperature coefficient shows a positive exponent (XI 05) whichis n6t'consisten t; with the equation. ,Please correCt. ' 'e RPI Response - Figure will be changed (see revised Figure 13.2 below) in the next SAR update.:
c.. On Figure 13.3 the finalexponent (-4) of the equation~is missing, possibly a photocopying artifact. Please correct...,,
RPI Response - Figure will be changed (see revised Figure 13.3 below) in the next SARI update.
Page 25

RAI Response Revised Fig. 13.2:
A LEU CORE A SOLID CORE ISOTHERMAL TEMPERATURE COEFFICIENT (for 421 pin core, 0595-inch pitch)
L~.
Pf~t         a
     -rt-g -r, Tt Wraw- C,..YF
      -26810, 7r SData point derived from LEOPARD mid DIFXY computer code analysis Data point plotted from quadratic fit tb 7 computer generated coefficients Page 26

RAI Response Revised Fig. 13.3:
                                          , LEUCOREB
                                           -ANNULAR CORE ISOTHERMAL TEMPERATURE COEFFICIENT (for 424 pin core, 0.585-Inch pitch)
                            .   ('e.)          I16' WT1-!rS31rLCsW't jiC                                  4 l.VY213cIC a                    'a rr~      -                     -                   -
f TO. 3z,'&              a'lr.
 -ftc/a-Ir Data point derived from LEOPARD and DIFXY
                   *'computer code analysis             -
* Data point plotted from quadratic fit to co~nputergenerated coefficients
  -3b 00--
V..
Page 27

RAI Response Questions related to the Technical Specifications General In accordance with 10 CFR 50.36, provide proposed Technical Specifications (TSs). The
,proposed TSs should be in conformance with ANSI/ANS-15.1-1990, "American National Standard for The Development of Technical Specifications 1for Research Reactors," as.
appropriate. The standard provides valuable, guidance &#xb6;in' the developmrent, of the TSs
-such that they meet the requirements of 10 CFR 50.36. Each individual change in the proposed TSs from the current TSs incorporated in, Facility Licehse..No. CX-22 (current TSs) should be cited. Substantive changes should be justified with analysis or discussion, as appropriate. In addition, each TS editorial change should be described in your response. Change citations and the accompanying justifications -and 'descriptions should not appear in the proposed TSs8 The prioposed:TSs shallbe reviewed and approved by the Nuclear Safety Review Board in accordance with the Administrative Controls required by the current TS 6.1.5.3, "Review 'and Approval Function.";
Pursuant to 10 CFR 50.36(b), the. technical specifications will be derived from the analyses and evaluations included in. the safety analysis report (SAR). Many of the following RAIs request you to provide reference to analysis in the SAR as basis, justification of the TSs. This may: be accomplished by referencing analysis already contained in the SAR, providing replacement SAR pages that contain the analysis, or by providing a separate analysis, discussion, and/or reference. In the latter case, the staff may incorporate ;that response in!its!,Safety:Evaluation Report by: reference, and you may provide replacement pages for your SAR at a later time.                              .
Pursuant to 10 CFR 50.36(a), summary statementdof the bases or reasdnsfor such specifications, -other than those;covering adminisfrative controls, 'shall also.'be included in the proposed specifications, but shall not become part of the technical specifications.
Pursuant to 10 CFR 50.36(b), the Commission may include such additional TS as the Commission finds .appropriate,,and the approved TSs and any additional TSs will be incorporated into the renewed license:.              '       '      .    '
The following is, a list of specific sections of the.proposed TSs submitted as Appendix A of the "RPI, Reactor Critical, Facility Relicensing Report,"'.with ,your~app*ication dated November 19, 2002, that require clarification:oradditional information...<':            ,
General                       v'             '         ' ' ,   '           .  ''
The proposed TSs should be included as a separate attachment to your response to this letter.
The proposed TSs should not have the heading, "RPI Reactor Critical Facility Relicensing Report, 12/2002," that appears on each page of the proposed TSs .submitted.
November 19, 2002.
Page 28

RAI Response The proposed TSs should have a title page and table of contents, similar to those contained in the current TSs.
122Format 1.2 &#xfd;Section, 1.2 of the&#xfd;proposed TS references 4ANSI/ANS 15.1. Update this reference to include the appropriate revision date and ensure -that alltreferences toANSI/ANS.15.1 that appear in the TS.s are.to the same revision; of the standard. (See TS RAI 1.3.X)
RPI Response'- ANS 15.1-2007 is theconsistent referen'ce.
1.3 Definitions, 1 .3 The terMs:"known, core'"7 and "unknownoruntested core",,appear in the TSs, butare not defined. Provide definitions of these terms.: (See TS RAI 4.1 :(A))
RPI Response - These definitions havebeen added:
1.3.D The definition in the current:TSs contains referencesto EuO3 in a.stainless;steel cermet, stainlesssteel, and an alloy of silver-cadmium-indium as possible-materials for the control rod absorber sections. The&#xfd; proposed, definition does not reference.these ,
materials. Confirm that these materials will not be utilized for the control rod absorber sections:. ,
RPI Response     - The only poison intended in the control, rods is boron. The definition has been revised.                                  I L; 1.3.0 (A);The definition ofreactor shutdown-is; circularin that; it icontains the phrase, "reactor is shutdowh by at least J.00$J':-Revise the definitionito-eliminate the circularity.
RPI Response - The definition has been revised to be consistent with ANS-15.1-2007.
1.3.0 (B) The; definition does not acd0untfor all possible states of the reactor: For example, if the core contains 50% of the fuel pins required for criticality and a control*.
rod is manually withdrawn (e.g., for maintenance or testing), the reactor is neither secured, nor shutdown, nor bperating, Explain any' formal controls in place to preclude the&#xfd; reactor'being in-an, undefined state,; orrevise'this definition to eliminate-the possibility that the reactor could.be in an undefined state.:.
RPI Response - Definitions are now provided for reactor operating, reactor shutdown and reactor secured. RPI believes these definitions are mutually exclusive and all inclusive.
1.3.0 (C) Consider adding a separate definition for "Reactor Operating," instead of including it in the definition of "Reactor Shutdown."
RPI Response- Done, see item above..
Page 29

RAI Response 1.3.P The proposed definition specifies the maximum permissible distance and travel time for the Licensed Senior Operator (LSO) on call as 30 miles or 60 minutes. The current definition specifies the maximum permissible distance and travel time for the LSO on call as 15 miles or 30 minutes. Provide justification for the increases in the permissible distance and travel time. (See SAR.RAI 12.6)
RPI Response - A justification is provided in the response to RAI 12.6.
1.3.T Provide justification that the restraining forces that hold the fuel pins in the.reactor core will be adequate to restrain any secured experiment. Alternately, revise the requirements for the magnitudes ofrestraining forces needed to ensure that secured experiments will not become unsecured during normal operation and, credible accidents.
RPI Response - The definition~has been changed to agree with ANS-15.1-2007.
1.3.V The SAR and TSs refer to both shutdown reactivity and shutdown margin as though the two terms are interchangeable.. Provide a definition of one of the terms and use that term consistently throughout the SAR and TSs. (See SAR RAI 13.5 and TS RAI 3.2 (D))
RPI Response - Both definitions have been revised to be consistent with ANS- 15.1-2007.
1.3.X The definition references.standard'ANSI/ANS 15.1 (1982). Ensure that revision of the standard is the revision referenced throughout the TSs. (See TS RAI 1.2)
RPI Response - See response to RAI 1.2 above.
2.0 Safety Limits and.Limiting Safety System setting ,.,
2.1 (A) The SAR contains nodiscussion of the technical basis for the safety limit.
Provide discussion and analysis of the technical basis for ihe safety limit.
RPI Response - The original safety limitfhas been deIeted.dNo'safety limniit is provided in the new Technical Specifications. This is adequate according to ANS-15.1-2007.
2.1 (B) 10 CFR 50.36c(1)0)(A) requires'safety limits "upon importanf process variables that are found to be necessary to reagonably protect. the integrity of certain of the physical barriers that guard against the uncontrolledrelease 'of radioactivity." TS 2.1 does not adequately address protection of the fuel cladding integrity. Provide analysis that shows that no material degradation of the fuel cladding will occur if the fuel pellet temperature is limited to 2000 *C. Otherwise, revise the safety limit and provide analysis or discussion that shows the new safety limit will reasonably protect' the integrity of the fuel and the cladding.
RPI Response - The accident analysis of SAR, Chapter 13, shows that no significant reactor temperature increase occurs for the accident conditions postulated to be the most Page 30

RAI Response severe. Thus this' accident does nro"t challenge"ihe stainless steel exterior cladding of the fuel pinrs, As perrmitted by ANS-l5.1*2007, no safety imit has been given in the Technical Specoificat0ins.. The accident conditidris stipulate a maximum excess reactivity and this is restricted by naking excess reacti ity a limiting coditiojY for operation. In conjunction with the limiting safety system Settings, th anaiyzed accident does not damage the fuel pin.
2.1 (C) The reference to W.A. Duckworth, ed., "Physical Properties of Uranium Dioxide," Uranium Dioxide: Properties and Nuclear Application's (Washington; D.C.:
Naval Reactors, Division0of Reactdr'Develonrrient) , 1961, pp. 173-228, that appears in the current TS does riot appear in the propdsed TS. Provida reference to this'docum-ent orf reference to analy'si'S'ithe SAR that supports the bastis-forTS 2.1.
RPI Response Since the safety limit does not' involve fu'el emprature,! there is no need' for this reference.
2.2( A) I0 CFR 5 .36c(1)(ii)(A) requires that the'lmiting safety system setting must be so chosen that automatic protective action 'will cofrect any abnoriiaal situation before a safety limit is exceeded. Provide reference to analysis in the SAR that demonstrates :the limiting safety system settings for reactor power and reactor period will not result in the safety linit being exceeded.-                         '.      ''
RPI Response - The SAR, 'Chapter 13, sedtioiit13.2, degcribes the accident conditions and the consequences.
2.2 (B) The bases of TS 2.2 refer to "energ 'ddposition,.' ".entli'Apy'rise'" and["power increase," whereas the safety limit is specified on the fuel pellet temperature. Provide reference to analysis in the SAR that telates 'thetliree abov&-me1/2nnioied terhms'to fuel pellet temperature, or revise the.bases of TS 2.2 to use temperature-related terminology with reference t6o uppjrng analysi*inthe    h SAR.I RPI Response - These terms have been rempved.
2.2 (C) Given that TS 3.2.7 requires a mimmum of 2' oUnts per seconidoin the start-up channel and TS 3.2.9 requires an interlock blocking rod withdrawal when neutron flux is less than 2 counts per secdndI'emove the limiiting'safety system setting foir minrimum flux level 'and lhe associated basis from TS 2.2. ,.
RPIR'sonse - Deleted from TS22.21'.                             .
3.0 Limiting Coidiions for Oper6tion 3.1 Section 13.2 of the SAR lists the initial temperature of the reactor coolant as 20 C.
10 CFR 50.36c(2)(ii)(B) requires a technical specification limiting condition for operdtion (LECO)bn "a process variable, design feature, or operating restrictioh that is an initial condiiionof a design basis a ccident or transient analysis.." Accordingly, propose Page 31

RAI Response a technical, specification for the maximu~m reactor, coolant temperature. Include a basis that references analysis contained in the SAR.
RPI Response - The moderator temperature was given as 2o C in the SAR for, reference purposes only. The temperature of the, moderator was not an initial condition dr variable in the accident analysis, and therefore no TS amendment is needed to specify an LCO for maximum coolant temperature.
3.1.2 Given that TS 3.1.3 allows reactor operation at temperatures 50 F and above, set limits on the void coefficient of reactivity in the temperature range from 50 *F'to 100 F, or provide justification for not ,doings9.so RPI Resp~onse - Boiling of reactor moderator is not a credible scenario' given the reactor power (< 100 W) and-vol-ume (2000 gallons). The most severe accident does not demonstrate that voids form, nor is anyvoid coefficient assumed in the analysis.
Therefore the void. coefficient of,reactivity is not limited by the technical specifications.
3.2 (A) 'Updatetthe reference to the "Hazards Summary Report" to reflect the current safety analysis document..
RPI Response - Technical Specifications were changed.
3.2 (B),-Footnote (a) to Table .lindicates that the "Log Count Rate".safety channel, may be bypassed when linear power channels are reading greater than 3x 10-10 amps. Provide the count rate or power leve!, that correspondsto, 3x10-10. amps.
RPI Response -- The footnote has been, removed' With the current. instrumentation, the interlock can't be bypassed.
3.2 (C) Update Table 2 of proposed TS: 3.2 -to ,refleet Amendment No.t11 to Ficility License CX.22: dated 'September 7,' 2004,;whichapproved removal :of the interlock, "Failure of 400 Cycle Synchro Power Supplty" RPI Response - Table 2 of the current Technical Specificatdn is correct'with Amendment tI.,  1..:
3.2 (D) Table 13.2 of the SAR lists the value of theshutdown margin used in the accident analysis as >0.02. f0 CFR 50.36c(2)(ii)(B) requires a LCO on a process variable, design feature,,or. operating ,restriction that is anlinitial condition of aqdesign basis' accident or. transient =nalysis.. ..* Accordingly, .propose &#xfd;a technical specification for; the shutdown margin. Jncludea .basis thatreferences analysis contained inthe SAR. ;(See.
SAR RAI 13.5 and .TS RAt 1:.3.V)               .
RPI Response - The proposed TS 'requires a shutdown rTactivity             > $1 'with ,atl four control rods fully inserted. See Section 312. The accident'analysis in theSARh i'*s been updated to use this value.
Page 32

RAI Response 3.2.1 (A) Provide' discussio6n and/or afialysis in the SAR of the technical base's for the' core excess reactivity and the maximum reactivity worth of a clean fuel pin. Provide reference to that discussion and/or analysis in the bases for TS 3.2.1 RPI Response - pagrajph Will be added to the SA9, section 13.2, on the next update stating that maximum reactivity worth of a clean fuel pin aie setto prohibit the possibility of exceeding the excess reactivity limitation. TS 3.2.1 has been amended to reference that section.                  !~   ~' ,.,   -?    -{*:, ,  "'  '" .          .
3.2 1 (B) Table 13.2 O6V     the SAR gives the' reactivity worth of a standard fuel assembly as
<0.039, which does not appear to be consistent with the maximum reactivity worth of 0.20$ specified by proposed TS. 3.2.1. Explain the apparent discrepancy or update Table 13.2 of the SAR to be consistefit with the probps6d TS 3d.2.1. (See S.. RAI 135)-
RPI Response- The confusion here is due to tooissio   *ns'' n of a $. See'the revised Table 11*2 entry. provided with the response to SAR RAT 13.5. The maximum allowable reactivity Worth of a fuel pin is $0.20, as written in TS 3.2.1. The SAR will be updated..
3.2.3 Provide discussion and/or analysis in the SAR of the technical bases for the maximum control rod reactivity rate. Provide reference to that discussion and/or analysis in the bases for TS 3.2.3.                 '"      "'"                             ".
RPI Response - The limit on maximium c6ntroi rod readcfivity'ate has been deleted as unnecessary. The response to RAI 4:4 discusses why a high'rate is not credible. Further, the analyzed accident does niot initiate from control rod motion, but'from a'step insertion of reactivity, far greater than, would .occur from control rod motion. Therefore, no limit on control rod reactivity insertion rate is needed.
3.2.4 Clarify whether the, magnet.release time of-50 mitliseconds includes the safety system response time, i.e., the time required for interruption of power to amagnet once a.
measured value reaches the safety system setting.,, If not, revise TS 3.1.4 to include the safety system response time and provide reference to appropriate analysis in the SAR.
RPT Response - The accident analysis in the SAR will be updated to include the safety system response time. The magnet release time of 50 ins does not include the safety system response time. Th6 Technnical Specifieation has been revised.
Note that the, accident scenario of SAR, Chapter.13, provides 1.5 seconds for control rod scram,, but allows a full 5 seconds., before reactivity is inserted. The revised Technical Specification-allows 900 millisecondsfor rod drop time, initiated by a manual scramo signal, and 600 milliseconds for instrument response. Magnet release is. measuredas part of the rod drop time..The revised Technical Specification does not require a separate measureimnentif magnet release time.             '                                    "
3.2.5 The basis for TS 3.2.5 states that "the requirement that negative reactivity be introduced in less than one minute following activation of the scram is established to minimize the consequences of any potential power transients." The SAR does not mention any power transients the consequences of which would be minimized by the Page 33

RAI Response auxiliary reactor scram (moderator-reflector water dump), nor does the SAR explain the technical basis for the requirement that negative reactivity be added within one minute of activation of the auxiliary scram. As written, the SARprovides inadequate justification for considering the auxiliary reactor scram a safety feature, and therefore, according to 10 CFR 50.36c(2)(i), a LCO should not be placed~on the auxiliary scram. Provide discussion and analysis in the SAR of the technical basis for the safety function of the auxiliary reactor scram, including quantitative analysis of the requirement that negative reactivity be added within one minute of its activation, or remove TS 3.2.5 and its associated bases from TS 3.2. In addition, modify TS 3.2.8 as appropriate.
RPI Response.- The auxiliary scram function of the moderator dump. valve is not included in the accident analysis in the SAR.. However, it is easy to maintain, and provides protection in the extremely unlikely event of multiple stuck control rods.
Therefore, it is being maintained as a required safety system in the TS. The requirement that negative reactivity be added within one minute of the dump valve. opening provides a means of periodically assessing the viability of the.dump valve as a secondary scram mechanism.
3.2.6 (A) The current TS 3.2 contains a basis for TS 3.2.6 that states, "the normal moderator-reflector water level is established not-greater than 10 inches above the top grid of the core..." The proposed TS 3.2 does not contain a basis for proposedTS 3.2.6.
Provide reference tO analysis or discussion in the SAR of the technical basis for establishing the moderator-reflector water level not greater than 10 inches, above the top grid of the core.      -          ,
RPI Response - A paragraph has been added.o the TS explaining the basis for-the limit on water height. The requireiment that thfh& teri level' be-no greater than' 10 inches above..
the top grid of the core is a means of ensuring that the time taken to insert negative reactivity via the secondary scram is not greater than the time measured during surveillances. 'The SAR will'be revised on the next update.
3.2.6 (B) Justify' not specifying a limit on the minimum moderator-reflector water level, or include a LCO on minimum moderator-reflector water level and an associated, basis with appropriate reference to discussion or analysis in the SAR.
RPI Response - The reactor is under-moderated andoperates at low thermal power (< 100 W), and therefore having a low moderator height is not a reactor safety concern.
Measurements with water level lowered are performed with a specific experimental procedure.
3.2.8 See TS RAI 3.2.5.               "
RPI Response    -  See the response to 3.2.8.
3.2.9'(A) Table 2 provides insufficient information about the interlocks that prevent rod withdrawal. Include the appropriate symbols (.i.e., <, >, and/or =) for "ReactoiPeriod 15 Page 34

RAI Response in sec" and "Neutroh Flux '2 cps," such thiaithe iiiteilbcks are consistent with the analysis theSAR. Include th frillure coi[difibo'6r'coniditi6ns 'for "Failure ofLine Voltage to Recorders" (e:g:, lihe vdltag'l~ss'thdm "X"*!6lis,)2            ..
RPI Response`-- The table has been 'upd.tedo           c....d 0r spec ific inforation interlocks:
Line Voltage to-Recorder< O00 V.
Reactor Period <---!'15 sc6nds         "
Source Rane.&#xfd;Counts <=2 counts 'ie'rsecond 3.2.9 (B) Table 2 'of the'cuirren'*TS 3:2 specifies the interl1ck "Water Level in Reactor Tank 10+/-1" Above-Core Top Grid." Table 2'of'the proposed'TS 3.2 does not specify that interlock. Providi justificatioh fo.r h6t includingthat intgrlo'ck in tie pfoposedTS 3.2.
RPI 'Response': A.gatige located in the trctor conf'otl'0r6odm provides indication of the water height. :The 'Startup ProCedures' reciied that water height in the reactor tank be visually verified after the fill pump is turned OFF, and prior to the operation ofthe reactor, to verify that the water level is at the desired height. Therefore, it is not credible for the water height to exceed' ti''abdve the cd66grid Without the knowledg'e of the operator. An interlock is not necessary" Fufther, a lter Waterlevel 'does not exacerbate the analyzed 'accident.              ',      "-,   .         '  .        '
3.2. 10 (A) Sections 1.2, 3.1, 4.1, 4.4,"and Table'41 of the `SXAR make references to an administratively-imposed maximum thermal power level of 15 watts and other *perating thermal power levels below 100 watts,. Confirm that the safety conclusions presented in the-SAR do hlt take ctedit fo?'ipo&r le&el6'lesstan'i00 Watts'specified as              Ts 3 .2 .16 :                '       .'    .  ..             .   ...
RPI Response'- The safety conclusions mithe. SAR do not take credit for a thermal power level of less than 100 watts. The'admisttatfive i it of 15waIts is i place because the reactor experiments do not require a higher power level, and the lower power limit helps.
keep exp~su're to persoinelhandlinig fu and workiiin in th6control room and control room hailway'ALCARA.                                                   "
3.2.10 (B) Provide a basis for the specification that integrated thermal power for any consecutive 365 days shail not exceed 200 kilowatt-hoUrs. Provide reference to analysis in the SAR that supports the bsis            '
RPI Response - The integrated annual thermal power limit of 200 kWh is not used in the safety analysis. This limit is used to ensure that the annual public exposure does not exceed expectations due to abnormal power and duration of operation for a given year.
Note that the proposed Technical Specification has changed this limit.
3.3.1 TS 3.3.1.c uses the phrase, "whenever the reactor is to be operated." This phrase is not defined in the TS and appears redundfiht to'the eneral applicability of TS 3A.
Reword TS 3.3. .c to clarify whether particulate monitoring is required whenever the Page 35

RAI Response reactor is not secured, or whenever the reactor is not secured and not shut down. (See TS RAI 1.3.0 (C))
RPI Response - TS 3.3.lx.c has been amended to remove the requirement for particulate justification..d air monitoring. See RAIJ7.6 for 3.3.2 Include the'minimum inventory and types of pQrtable survey, instruments required by TS 3.3.2, or provide justificatiQn for not including this information in.TS 3'.3,.2.
RPI Response - This section will be updated to state that ata minimum:.:,
During A~ormal operation, a calibrated and operational portable~s1urvey meter capable ofme'asuring ambient raadition exposure will be&available
   ',During fuel: loading or'unloading; or during any experiments involving the addition or removal of material fromh the c6re (activation' foils, edtc.)a thliii-win'dow GM'detector will be available-to check for tpersonnel or area contamination. .         ..     .'   -
3.4 The bases for TS 3.4.8 andTS 3.4.9.coptain outdated references to10 CFR,20.101, 10 CFR 20.103, 10 CFR 20.105, and 10 CFR 20.106. Update these references.
RPI Response - See the Techniial Specifications.
3.4.3 TS 3.4 does not contain a basis fri the reactivity worth or allowed frequency of moveable experiment which may be oscillated in the core. Provide a basis for TS 3.4.3 that references analysis in the SAR.
RPI Response - The limit on the maximum reactivity insertion rate due to an oscillating experiment is not a safety limit, and therefore is no~t treated in the SAR. This limit is, meant to maintain controllability of the reactor during the performance of experiments.
The bases has been revised.
3.4.5 TS 3.4.5 appears contradictory to the., equirements of-TS 3.4.8 and TS 3.4.9 regarding materials that may produce airborne radioactivity. Clarify the intent of TS 3.4.5 as it applies to experiments that are not encapsulated, singly-encapsulated experiments, and doubly-encapsulated experiments.
RPI Response - Thie specifi&ation gitvenby TS 3A.4.5 is a blanket statement,that pertains to all experimentSWhether sin~gly)-, doubly-,.or un-encapsulated. TS 3.4.8 establishs, specific liiiits for encapsulated experiments. For clagity, 3.8T5 of the revised .gThnical Specification has been amended to read the following:
Experiments shall not contain materials which, can cause a violent chemical reaction. Unencapsulated experiments shall, not-contain a material that may.
produce significant airborne rradioactivity. Encapsulated experiments may
:contain materials that canl cause a minor release of airbornie radioactivity, subject tothelimits inTechnical Specifidations 3.8.81' Page 36

RAI Response 3.4.8 (A) The exposure time for persons in unrestricted areas (2 hours)!must be consistent with the ability and any plans RPI has in place to control occupancy of unrestricted areas, i.e., public evacuation plans and procedures. If RPI does not have approved plans and procedures for controlling oculanchy in unrestricted areas, the exposure time for persons in unrestricted areas should be based on the maximum possible exposure time for a release from.the particular experiment and the reactor building (e.g.,
plume passage time). Provide justification of the use'of a 2-hour expos ure time, or revise the TS to account for the maximum' possible exposure time.
RPI Response -RPIdoes not see a need to specify a: universal limit for maximum activity must  review allmaterial of radioactive            that can experiments      bfe placed prort                 core for any experiment. The NSRB m in' the .....
prior to implementation. This review considers failure of any experiment containing radioactive materialland ensures thati failure will not compromise regulatory exposure limits. There isalso no need to treat singly encapsulated and doubly encapsulated experiments differently, so TS 3.4.8 and TS 3.4.9 (renumbered as 3.8.8 in the revised Technical Specification) have been combined, into asngle condition.
3.4.8 (B) Provide a discussion of the method used to ensure compliance with the requirements of TS 3.4.8. Include the methods and assumptions used to calculate doses to persons in the restricted area and unrestricted area.
RPI Response - see response to 3.4.8(A) "above 3.4.9 (A) See TS RAI 3.4.8 (A)
RPI Response    -  see response'to 3.4.8(A) above          .                   .
3.4.9 (B) See TS RAI 3.4.8 (B)
RP[ Response '-seeresponse to 3.4.8(A) above
: 4. Surveillance Requirements' 4.0 Specify surveillance methods, requirements, and acceptance criteria to ensure monitoring of the"fuel integrity and preclude'theuse of damag6d (e.g., corroded, bowed, leaky, etc') fuel pins, include abasis that refernces or sumnmarizes discussion in the SAR.: Otherwise" provide justification forrhot requiring surveillance of the fuel pins.
RPI Response - RPI judges that no Technical Specification surveillance is necessary.
Discussion will be included in the SAR on the next update.
The fuel is stored dry and only wetted infrequently, usually once a week for a few hours.
The fuel pins do not show any sign of corrosion after. 20 years of operation: Pins are removed from service if they become bowed because the pins are then difficult to align with the upper and lower support plates. Frequent radiological surveys for loose surface contamination would detect a leaking fuel pin. RPI notes that twenty. years of use of these Page 37

RAI Response pins has never detected any leaking pins. If a pin' wete severely damaged. during handling, the event would be treated as a radiological casualty andsurveys. taken to determine if the pin were now leaking.
4.1 (A) TS 4.1 refers to an "unknown or previously untested core." The proposed TSs do not provide a definition'6r'the characteristics ofan unknown-or untested cbre. Provide a definition of an unknown or untested core. Revise the basis for TS 4.I to summarize or reference discussion or analysis in the SAR that addresses the specific qualitative and/or quantitative characteristics that differentiate an unknovn'bor untestd' core from a known core. (See TS RAI1.) Y"                       "           ..
RPI Response - The term known core is defined in the Technical Specification.
4.1 (B) The basis for TS 4.1 irfersto the initial testperiod of'thereactor. Provide-clarification as to WhIether the initial test period of the recactor'is the initial test period for any unknown or ffntested coreand revise the basis for TS 4.1'as appropriate.
RPI Response - ee the revsed Technic'             pecification.
* 4.1 (C) Provide reference to a'na ysis or-discussion in the SAR that describes the methods used to determine theiractr p&#xfd;atmeters specffied ifi TS. 4.1 during the initial testing of an unknown or untested core.. rncrude a discussion of safety precautiofis and controls.
RPI Response - See the revised Techfnical Specificaiion. The experimental procedures; developed for the facility are those also used to determine that the reactor parameters 'of a new core configuration. meet the requirements of Section 4.1. The SAR will be updated.
4.1.a The proposed TS contains the word "back," while the current TS contains the word "bank." Clarify           .''     .                  '1/2 4    .
RPI Response - The correct word is "bank". The typographical error has been corrected 4.1.d See SAR RAI f 3.5, TS RAI l'.3N, and TS RAI 3.2'(D)'
RPI Response       See the responses with the refereniced questions*.-
4.2 (A) 10 CFR 50.36c(3) requires surveillance requirements "to assure that the necessary quality of systems and components is maintained, that facility operations will be within safety limits, and that the limiting conditions for operation will be met." TS 4.2 does not specify surveillance requiremIents'io'support each teichni'c specifiCaltibn in TS 3.2,speci'fically TS 3.2.1,' TS 3.2.2, dnd'TS' 3.2:3. 'Prp1ose 'appropriate tiivei liifice requirements to verify each LCO in TS 3.2, or justify ornitting-surveillance requirements.
RPI'Response - See the revised Technical Specification            "
4.2 (B) The first paragraph of the bases states, "past performance of control rods and control rod drives and the moderator-reflector water fill and dump valve system have Page 38

RAI Response demonstrated :that testing semiannually is:adequiate to ensure compliance with Specification 312, Itenms: 3, 4, and 5:" Please clarify which surveillance requirement specified by TS 4.2 ensures compliance with TS 3.2, Item 3.
RPI Response - The statement has been corrected to refer only to itemns corresponding to the discussion of bases.
4.2 (C) The second paragraphof the bases States, ,redundancy of allsafety channels is provided..." Table 1 of TS 3.2 requires a minimum of 1 "log count rate" safety channel and 1 "log-N; period" safety channel. Clarify the apparent discrepancy.
RPI Response - The bases'are poorly worded and have been' revised to show how redundancy is. achieved.. There are three instrument channels that use ion chambers and all three, havea high currentscram. Qne of thes6e;,the 16g-N channei also has a fast period scram. The Log count rate chaninel is driven by a BF3 detector and piovides the rod outmotion interlock on low counts. Redundancy is claimed based on the ability to still provide a high current scram even if any.one of theion chamber channels fail. Note that the SAR analysis assumes the failed channel is fthe Log-N channel so both a high current and a fast rate scram are unavailaqle. The high current.scram from either linear power channel terminates the accident. The accidenhtis independent of power level so failure of thelow source counts instrument does niot 'increase the severity of the accident.
One could argue that the log count rate channel is nota safety channel, however it is listed here as a necessary channel for operation to, assist the operator as the reactor approaches criticality.,         .
4.2.5 Verify that the reference to TS 3.2.5 should nodt be TS 3.22.6 RI Response '-The reference has been changed to 3.2, Ifem 5. Renumibering occurred in this section.
Suvilac
                      .rps euieet              f .       t... .... ....      K 4.3 Propose surveillance requirements for the portable detection and survey instruments specified in TS 3.3.2, or.justify omitting surveillance requirements.
RPI Response - Periodic instrument calibration, requirements have been added to the Technical Specifications'.                                  '    "
: 5. Design Features,         . ..             ..
jirstparagrfJ of                        if 5.4-.1, The flst paragraph of this. secti'on statesithiat the reactor tank is a stainless steel'
        -,lined tank. .Section 4.3 :of the SAR states ih at fhe reactor tank is stainless steel.
        .Clarifythe apparent discrepancy.
RPI Response - This typographical error has,.been corrected. The tank is stainless steel, not lined with stainless steel.
Page 39

RAI Response 5.4.2  This section describes the core as consisting of all SPERT (F-i) fuel, or approximately half of SPERT (F-I) fuel with the remainder being low enriched uranium light water reactor type fuel of typical power reactor design and arrangement. If the intent is to maintain this capability; provideadditional information on this latter fuel such as :design parameters, qualification, and operating limits. Describe any special handling or storage considerations.
RPI Response - The text describing fuel other than SPERT (F-I) pins has been removed.
No other fuel is at the facility.
5.4.3  The first sentence of the second paragraph of-this section contains a typographical error in that the word "on" appears in~the-proposed TS where the word "one" appears in the current TS. Clarify.
RPI Response --The correct word is -"one". The typographical, error .has been corrected.
The correct reference is to5.4.4 vice 5.4.4.
5.6 (A) The core loading specifications described in this section are LCOs and should be properly formatted and placed in the 'appropriate sections of.TS 3. Otherwise, justify not making such a change.-,.
5.5   (B) Item 4 of the proposed TS contains a typographical error in the word "one" appears in the proposed TS where the word "oi"'.. appears, in the current TS.
Clarify.
RPI Response - The correct w6rd is "'on". The typographical errorhas been corrected.
The correct referenceisto 5:6.vice 5-5.
: 6. Administrative Controls RAIs pertaining to TS 6 are incorporated in the RAIs icoverifng Se'ction 12 of the SAR Page 40

Safety Analysis Report                                                         June 2008 12 Conduct of Operations 12.1:      Organization 12.11 'Structure - Responsibility for the safe operation of the reactor facility is vested within the Chain of command shown- in Figure 12t1.
Figure 12.1: Facility Organization Page 1

Safety Analysis Report                                                              June 2008 12.1.2 Responsibility 12.1.2.1 Level 1: The Dean, School of Engineering; is'responsible for the facility license: and appoints the, Chair, Nuclear. Safety Review Board.
12.1.2.2 !Level 2: The Facility Director is responsible for facility.
administration and safety. The Facility Director reports to the Chair, Mechanical, Aerospace and Nuciar Engineering, for administrative purposes.
12.1.2.3 Level 3: The Operations Superv            iis espnsible 'or the day-to-day operation of the facility and reports to the FacilitYIirecto        ,
12.1.2.4 Level 4: Licensed operators and senior operators are the operating staffand report to the Facility Director, for administrative purposes.
12.1.2.5 TThe RPI Radiation Safety Officer (RSO) who :is organizationally independent of the reactor bperatidns group shall'provide advice as required by the Facility Director and the Operations Supervisor in matters concerning radiological safety. The RSO also has interdiction responsibilityand authority.,
12.1.3 Staffing 12.1.3. f Theminimal staffing when the reactor is not shutdown shall be:
12.1.3.1.1       A senior reactor operator licensed pursuant to 10 CFR 55 present at the controls.
12.1.3.1.2       One other person in the control room certified by the Operations Supervispr as qualified to activate manual scram and initiate emergency procedures. A second senior reactor operator or a reactor operator licensed pursuant to 10 CFR 55 fulfills this requirement.
12.1.3.1.3       A licensed senior. reactor operator~shallbe present or readily available on call. This is defined as being With,60 minutes normaltravel time, or 25 miles, whichever is more limiting. The time for the on,.call operatorto arrive is based on reasonable response to potential needs that can't be satisfied by phone. It is considered urilikel)y that a second operator -wouldactually need to a-rive that quickly since the reactor can be placed in thersafe'shutdown mode in less than a minute.
12J.3-2        The identity, pf and, method, for rapidly contacting the licensed senior operator on duty shall be known to the operator.,.
12.1.3.3: No staffing is; required when the reactor-is in secure ishutd6wn.
12.1.3.4 The minimal staffing When the -reactor is shutdown buf not secure shutdown shall be a senior reactor operator at the facility and, a second senior reactor opera*br present or readilyavailable 0n cili.                      .
12.1.3.5 A list of reactor. facility:personnel by name and telephone number shall be readily available in the control room for use by the operator. The list must include management personnel, radiation protection personnel and other RCF Staff.
1.2.1.3.6 , Events requiring the direction of the Operations' Supervisor:
12.1.3.6.41     Allfuel:or control'rod relocations within the'reactor core.
Page 2,

;*Safety Analysis Report                                                               -June 2008 12.1.3.6i2     Recovery from unplanned or unscheduled shutdown.
          .12.1.4&#xfd; Selection and Training of Personnel New react0r operators and, senior, reactor operatorsi are selected from interested students enrolled in classes that takeplac6at the RCF. The Operations Supervisor is responsiblefor the operator training assisted ly other RCF staff.: The Operator Requalificatibh Prograr ineets the regulatidns in 10 CFR 55.."
12.1.5 Radiation Safety              .                            -
Radiation safety aspects of routine facility operation.aretypically performed by members of the RCF staff who receive training from the RSO to perform those tasks.
Thus radiation surveys tolverify: normal radiation levels during reactor operation, fuel handling, ortexperiments 'will be conducted by, the RCF Staff 'The RSO is available for assistance ifneeded. The RSO. also conducts periodic contamination surveys and maintains and monitors personnel exposure records.
12.2 , Review and Audit Activities The Nuclear Safety Review Board (NSRB) provides independent ieview and 'audits facility activities. The Dean, School of Engineering, appoints the NSRB Chair.
Some members of the NSRB are appointed by virtue of their position, the Facility Director and RSO are examples.of this. Other,members*of the NSRB are appointed by their management. The NSRB then reports to the Dean, School of Engineering.
The NSRB Charter provides additidnal details. .
12.3         Procedures                ,
12.3.1 Written operating procedures are usbdfor the following 12.3.1.1.1     Reactor Pre-Start.up,,
12.3.1.1.2     Reactor Operations, including conduct of experiments
                   -123.1.1.3,     Stirveillanices     ,                   .
2.31.14      Emergencies 123.1:1.5; t              RadiationPr'.tection' 12.3.2 Procedures are developed by the RCF Staff in response to a planned need for a new or revised procedure. Existing, procedures. are consulted. and revised if possible to meet the need for a new procedure. This process is supervised by the Operations &#xfd;Suitervisor. A proposed *he\ procedure' s'revieWed by the' Operations Supervisor and the Facility Director to deteimine the need for NSRB review and approval. Minor changes that do not affect, the:safe operation of the reactor may be approved for use by the Operations Supervisor. Procedures which do not meet these
      &trteriti are preseht'ed toapprova.
to'he NS"B for ap6l'.A           bVdP Apptrovedpr~cedures     are put into use after updating the' list df approved piroedures'. Thfsilist inform's the operators which procedures, by name and version, are approved for use. The list is updated and S.approyed by the Operations' Supervisor.and is' posted at or near the reactor operating console.,                     -           '             ,
1213.3 'Radiation protedcion procedures are maintained by the Office of Radiation and :Nuclear Safety, with the approval of the, Radiation and Nuclear Safety Committee. The Radiation Safety Manual addresses the Program, Policy, and Orgahization of the radiation safety program, the Radiation Safety Training program, Page 3
:,Safety Analysis Report                                                                 *June 2008 radioactive waste management, dosimetry and radiation monitoring; instrumentation calibration' and the ALARA program. The, Radiation, Safety Officer ensures that the Radiation SafetyManual addresses each. of the recommendations~in ANSI/ANS-15.11-1993, and distributing updates to the Critical Facility Director and Nuclear Safety Review Board..
12.4          Required Actions
      , 12.4.1 Action to be taken in Case of Safetylmit Violations - No action steps are provided for this since there'is no'identified safety 'limit. Safety'limitsb are not required for reactors' without engineered"safet;y' syste'ihs provided that the accident analysis shows that there Iis no damage to the furlcdding. Chapter'13 analyzes the potential accidents to theRCF and concludes that no fuel clad daniage will occur for the mOst'sexere'accident.        ..            , , .
12.4.2 'Action to be Taken in the Event of an Reportable Occurrence 12.4.2.1 The reactor shall be shut down. ..Operations shall not be resumed unless authorized by the. Chair,, NSRB.,
12.4.2..2 Occurrencie shall be reported to'the Facility Director or designated Salternate, the NSRB and to -the'Nucleaf Regulatory Comm'ission ntot*later than the
   'following workin'g.da-y by telephon&'and-confirmed in wrting to licdnsiing authorities, to be followed by a written report that describes the circumstances of the event within 14 days of the event.                 '
               "12.4.2.3    All such conditions,'including action taken toprevent or reduce the probability of a recurrence, shall be reviewed by the N$RB.; The NSRB shall concurwith corrective actions.             ,    ,    , .                     .
12.5           Reports Reports include-annual operating, reports that. describe the. activities for the previous year and"non-periodoicreports that describe impqrtant changes in the, facility or.
facility management. NUREG-1537,.Part 1.,.and ANS-15.-2,007 pyovide: guidance for the details of the information that should be reported. This guidance has been incorporated in the Technical Specifications.i,. !.-.'
12.6           Rec6rd Facility records are required to be maintained for specific periods of time depending upon the type of record. NUREG-1537, Part 1, and ANS-15.1-2007 provide guidance for the details of the retention time. This guidance has, been, incorporated in the Technical Specifications..
12.7          Emergency Planning 12.7.1 The RCF Emergency Plan describes the Critical Facility emergency.
organization and includes the responsibilities &#xfd;nd authority With :a line 6f succession for key members of the emergency organization' The emergency organization described in the plan ensures that emergency management will be availabl 't6 meet any foreseeable emergency at the research reactor. Additionally, the plan describes Page 4

,Safety Analysis Report                                                                June' 2008 the criteria for the termination of anemergency, authorization for reentry, and establishes limits of exposure to radiation in excess of normal occupational limits for emergency&#xfd; teamt members for life saving, andcorrective actions to. mitigate the consequences of an accident.
12.7.2 Two emergency classes are described for the Critical Facility. These classes are based upon credible accidents associated with the reactor operations and other emergency situations that are non-reactor related but-could affect routine reactor operations. The emergpncy. classes are.Personnel Emergency and Emergency Alert. Each class is associated with specifi Emergency Action Levels (EALs) for activating the ,emergency organization and initiating protective actions appropriate for the emergency event in proces.s. The Emergency Planning Zone, (EPZ) is the area within the Critic'ail Facility building. Predetermined protective actions for,the EPZ include radiation surveys to locate areas and levels of radioactive contamination, personnel evacuation should this become necessary and personnel. accountability.
12.7.3 The emergency facilities and equipment available for emergency response include a designated'Emergency Supp6rt Center,. radilogical monfitoring systems, instruments and laboratory facilities for continually assessing the course of an accident, first aid-and medical facilities and communications equipment. The provisions for,maintaining emergency preparedness include, programs for training, retraining, drills, plan review and updates, and.eqpipment inventory and calibrations.
12.8        Security Planning The RCF has established and maintains a program to protect the reactor and fuel and tolensure its Security. -Both the :physi'cals ecuiiity plan and the staff's evaluation are withheld, from publiic disclosure6 under 10 CFR23790(d1)(i) and 10 CFR 9.5(a)(4).
The current Security Plan was last revised in 2006.
12.9        Quality Assurance Quality, Assurance is achiedVdvia extehsive dcunrithation and periodic interaction with the Nuelear Safety Review Board (NSRB). Alilopeiatioris and experiments.
must 'follow writtenprocedulres that have been approved by the NSRB..
12.10       Operator Training and Requalificatio"n.
Operator training and requalification programs have been approved by the Nuclear Regulatory Commission.
12.11       StartupPlan A,startup plan isi not necessary for facility license.feneWal:. The facility is not undergoing any changes that would require such a plan'.
12.12       Environmental Reports The facility has existed upito the present without having any significant effect on the environment. No future changes to the facility are anticipated that would result in an nr*ea
             ; effe&#xfd;t increase'dd'    c oon tthe eenvironment.
                                    'f-        T1 facility has no off-site environmental:..
The Page 5

Safety Analysis Report                                                         June 2008 monitoring requirements. The annual operating report includes data on facility discharges and radiation monitoring data from site exclusion boundary dosimetry.
Page 6

TECHNICAL SPECIFICATIONS CRITICAL EXPERIMENTS FACILITY RENSSELAER POLYTECHNIC INSTITUTE July 2008 Approved:
Dr. Michael Podowski, Chair Nuclear Safety Review Board

This page is intentionally blank j
I, I.
I ;
4        -  0)

Technical Specifications Technical Specifications
: 1.       INTRODUCTION 1.1      Scope The following constitute the Technical Specifications for the RPI Critical Experiments Facility (RCF), as required by 10 CFR 50.36.
1.2      Application Content and section numbering are in accordance with section 1.2.2 of ANS-5.1-2007.
1.3      Definitions certificate or charter: See license.
certified: See licensed.
Class A reactor operator: See senior reactor operator.
Class B reactor operator; See reactor operator.
channel: A channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.
channel calibration: A channel calibration is an adjustment of the channel such that its out-put corresponds with acceptable accuracy to known values of the parameter that the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.
channel check: A channel check is a qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same parameter.
channel test: A channel test is the introduction of a signal into the channel for verification that it is operable.
control rod: A control mechanism consisting of a stainless steel basket that houses two absorber sections, one above the other. These absorber sections contain boron in iron clad in stainless steel. All are of the same dimensions, nominally 2.6 inches square, with their poisons uniformly distributed. The absorbers, when fully inserted, shall extend above the top and to within one inch of the bottom of the active core.
core configuration: The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and control rods Occupying the core grid.
excess reactivity: Excess reactivity is that amount of reactivity that would exist if all reactivity control devices and movable experiments were moved to the maximum reactive condition from the point where the' reactor is exactly critical (kff = 1) at reference core conditions or at a specified set of conditions.
experiment: Any operation, hardware, or target (excluding devices such as detectors, foils, etc.) that is designed to investigate reactor characteristics or that is intended for irradiation within the reactor.
1

Technical Specifications facility-specific definitions: Facility-specific definitions are those definitions uniquetoa specific facility.       -
known core: A core configuration for which the power indicating instrumentktion has been calibrated. in accordance, with surveillance procedures and the following parameters have'been measured ,
(1) excess reactivity (2) shutdown reactivity, allirods. inserted and onexrod stuck in.the full out position (3) reactivity worth of.most reactive fuel pin license: The written authorization, by the responsible authority, for an individual"or organization to carry: out the duties and, responsibilities associated .,with~a,personnel    .
position, material, or facility requiring licensing.,;
licensed: See. licensee.-
licensee: An individual or'organization holding a license.
measured value: The measured value is the value of a parameter as it appeas on the outpgut of a channel.       '.
movable experiment; A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.      ....
owner or operator: Seelicensee.,
operable: Operable meaIns a compOnent or system is Ccapable of peiforming its intended function..
operating: Operating means a component or system is performiiig'its intendedfunction.
permit: Se6 license'             ,                            -.
protective action: Protective action is the initiation of a signal or the operatioIn of equipment withinf the reactor safety system in response to a parameter or condition of the reactor facility having reached a'specified limit.                   '
reactor operating: The reactor is operating whenever the'&#xfd;reactor tank contains moderator and any fuel;, and any control rodi is not on the bottom:...
reactor operator:' An individual whios licensed o.manipulate ih6. corftols of areactor.
reactor safety systems: Reactor safety systems are those systems, including their associated input channels, that are desgned to ihitiate autornatic reactorprotection or to provide information for initiation of manual protective actilon."
reactor secured: A reactor: is secured w'h.
: 1. Eitherthere is insufficient moderator available in the reactor to attain criticality,,
control rods are inserted, and the console keys are removed,
: 2. Or all fuel pins have been removed from the reactor.
2

Technical Specifications reactor shutdown:"The. reactor. is -shut down~if all control, rods are insertedand it-is -*
subcritical by at least one dollar in the reference core condition with the reactivity worth-of all installed experiments included,.
reactivity worth of an experiment: The reactivity worth of anexperiment is the value of the reactivity change that results from the experiment being inserted into or removed; from its intended position.
readily available on call: Anr operator is readily :available on call if within 60 minutes normal travel time and 25 miles of the facility-and-personnel at-the facility can readily contact the individual.              .....
reference; core condition:, The condition of the core when:it is at ambient temperature:
(cold) and the control rods are on the bottom.
research reactor: A research reactor is defined as a device designed to support a-self-sustaining neutron chain-reaction for research-, development, educational, training, or experimental purposes and that may have provisions for the production of radioisotopes.
research reactor facility:' Includes all areas within which the owner or operator directs authorized activities associated with the reactor.
reportable occurrences
: 1. Release of radioactivity from the facility above allowed limits,;
: 2. Discovery of loose surface contamination, excluding contamination-due to naturally occurring radionuclides such as radon daughters;.,
: 3. 'Operation with actual safety system setting less conservative than the limiting safety system settings;,.
: 4. Operation in violation of limiting conditions for operation unless prompt remedial action is taken;
    ,5. Any reactor safety system component malfunction that could render the safety .
system iiicapabli of p'erforming its intended function; _
: 6. An unanticipated or uncontrolled change in reactivity greater than 60 cents; or
: 7. An observed inadequacy in the implemnentation of administrative or procedural controls. such that the inadequacy causes or could have caused the existence or ,
deveivoment of an unsafe c6nditionvw ih regard to reactor operations responsible authority: A governmental or 0ther entity with the authonrity to issue licenses, charters; permits, or certificates.;-
review and approve: The reviewing group or persons &sect;shall carry out a review.of the matter in question and may either approve or disapprove it: before it can be implemented, the mat-ter' in question mustreceive approval-fro'm the reviewing group or personis.
safety channel: A channel in the reactor safety system.
scram time: Scram time is the elapsed time between'the initiation of a scram signal and indication that the control rod has bottomed.
secured experiment: A secured experiment is any experiment, experimental apparatus, 3

Technical Specifications or componenif~of an. experimient that is:&#xfd;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,&#xfd;buoyanf, 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.
secured shutdown: The"reactor is secured and the facility adminitative requirements are met for leaving the facility with no lidensed operators present.:
senior reactor operator: An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.,
shall, should, and may: The word "shall" is used to dehote*arequireinht; th6eword "should" is used to denote a recommendation; and the word "may" is used to denote permission, neither a requirement nor a recommendation.
shutdown margin': Shutdown mar gin is the minimUrm shutdown reactivity necessary to provide confidence that the reactor can be made subcritical'by means of-the control and safety systems starting from any permissible operating condition and with the most reactive rod .in the most reactiveposition, and the nonscramable rods in their most reactive positions and that.the reactor will remain ,subcritical without further operator action.
shutdown reactivity: The reactivity of the reactor at ambient conditions with'all control rods fully inserted, inciuding the Ireactivity of installed experiments.,
supervisory reactor operator: See senior reactor operator surveillance frequency: Unless otherwise stated in theg specifieations, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. In cases where the elapsed interval has exceeded 100%
of the specified interval, the next surveillance interval shall commence at the end of the original specified interval. Allowable" surveillance fi*t*rvals, as defined in"ANSI/ANS 15.1 (2007) shall-not exceed-the foiidwing:                         '
Annual (interval not t6 exceed 15'montrhis).
Semiannual (interval. not to, exceed seven and.one-half months),.
Prior to the first reactor startup of the day..         .
surveillance' intei-val -'The -siurveillance interval is tfie calenddr time between surveillance tests, checks, calibrations, ad examinations to be'performed uponah instrument or com ponent When itis,friitei'ed'to: be operable. ..               u true value: The true value is the actual value of a parameter.
unknown core: Any core configuration that is not a.knowny. core.
unscheduled shutdown: An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor af9&ty system, ope'ratbr6 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..                 ..          ..
4

Technical Specifications
: 2.      SAFETY LIMITS AND LIMITINGSAFETY SYSTEM SETTINGS.
2.1     Safety Limits   -None, Bases The Safety Analysis JReport (SAR), evaluates all- potential accidents and identifies an unplanned or uncontrolled reactivity addition as themos~t severe. Analysis of this type of accident has been performed for an addition of 60 cents and acceptable core performance was demonstrated. See SAR;Sectiori13.2.            "
2.2     LimitingSafety System Settings.
Applicability Applies to the settings to initiate protective action for instruments monitoring parameters associated with the reactor power limits..
Qi6jective..
To assure protective action before safety limits are exceeded,.,
Specification                                                .                 .
The limiting safety system settings on reactor power shall be as follows:
Maximum Power Level                            .100 watts Minimum Period                  .              ` 5 seconds Bases                                               t    .             .
The maximum power level trip setting.of 100 watts on Log Power and Period Channel 2 (PP2) correlates with the operating license limit., The, scram setpoint is used in the safety analysis with the assumption that initial po~wer, is at 1,00 watts indicated power.
The minimum 5-secohfdperiod is specified'so -that the automatic safety system channels have sufficient time to respond in the event of'a very rapid positive reactivity insertion.
Power increase. and energy deposition subsequent to scram initiation are thereby limited to well below, the identified safety limit, This scram is not used in the analysis of the' most severe accident since the analysis assumes that the, safety channel, with a, fast rate scram fails concurrent with the reactivity addition.
: 3.      LIMITING CONDITIONS FOR OPERATION .
3.1     Reactor Core Parameters....,
Applicab.ilily    .,         -.
These specifications apply to reactivity in the control rods plus the maximum, reactivity..
contained in movable experiments.
5

Technical Specifications Otjective The purpose of these specifications is to ensure that the reactor is operated within the range of parameters:that have been analyzed.
Specifications The excess reactivity of the reactor above cold, clean critical shall notbe greater than 0.60$.
Bases Excess reactivity must be limited to ensure any reactivity addition accident is restricted to one that has been analyzed ard shown to cause no coredamage.-The assumption in this, analyzed accident is a step insertion 0f 60 cents of reactivity above critical.
3.2     Reactor Control and Safety Systeis:.-                                 .
Applicability Applies to all methods of changing core reactivity available'to the reactor operator.
OQjective To assure that available shutdown reactivity is adequate and that positive reactivity insertion rates are within-those analyzed in the SAR.
Specifications
: 1.      The maximum reactivity worth of any clean fuel pin shall be 0.20$.
: 2.      There shall be a minimum of four operable control rods. The reactor shall be subcritical by more than 0.70$ with the, most reactive control rod fully withdrawn.
The~minimum shutdown reactivity with all f6ir' control rods inserted shall be
        $1.00.                          .               ...
: 3.      The total control rod drop time for each control rod from its fully withdrawn position to its fully inserted position shall be less than or equal to 1.5 ,seconds.
This time shall include *. maximun instrument response timr of 600 milliseconds.
Instrument'response may'be measured'separately 'from rod drop time if desired.. If the total time is measured and is less than required, then instrument response time need not be~separately measured' to determine if .the600 millisecorid tirri is met.
4.. i The auxliliary reactor scramh (ioderdt*-reflectorwater, dump) shall add negative reactivity within one minute of its activation.
: 5.      The normal moderator-reflector water level shall be established not greater than 10 inches above the top grid of'{the ce o "%re.*
: 6.      The minimum safety channels that shall be operating during the reactor operation are listed in Table 1.
: 7.      After a scram, the moderator' dump valve may be re-closed by a senior reactor operator,,if the pause of the scram is known, all control rods are verified to have 6

Technical Specifications scrammed and it is deemed wise to retain the moderator shielding in the reactor' tank.
: 8.        The interlocks that shall be operable during reactoroperations are listed in Table:
2.
: 9.        The thermal power level shall be contrqlled so as not to exceed..100 watts, and the integrated thtrmalpower ifor any consecutive 365 days shall not exceed 2 kilowatt-hours.
TABLE 1: Minimum Safety System Channels Minimum Reactor, Conditions., Ranhges.                 -Channel S,                             Functions
: Number Start-up:          2 cps       - 10 4
cps     Log CountRate.            1        . Minimum Flux Level Power:                0 10Y1amps            Linear Power              2            High Neutron Level
                                                                                     - Scram 10-3         10-14   amps Log-N; Period                             High Neutron, Level
              +9991-        -999 seconds                                               and Period Scram Manual Scram(a)           2            Reactor Scram Building Power                         Reactor Scram
:* :    ::,:      ;, Reactor S r m b):Door ,. .      : t :'.  .. *' R
:. eactor Scram (a) The mianu'alscrar shall consist of a regular manualscram at the console and a manual electric switch which shall disconnect the el6ctricalpower of the facility from the reacto'i,' causing a~loss ofpdwer scram.
(b) The reactor door scram may be bypassed during maintenance checks and radiation surveys with the specific permission of the Operations Supervisor provided that no other scram chariinfs are bypase..                          '....
TABLE 2: Interlocks' Interlocks                                                        Action if Interlock Not Satisfied t
* rr           '  I                          n Reactor Console Keys &#xfd;2_)                 un                      Keactor cram Reactor Period <15 sec                                            Prevents Control Rod Withdrawal Neutron Flux <2 cps                                               Prevents Control Rod Withdiawal 7.

Technical Specifications Line Voltage to Recorders < 100 V                       Prevents Control Rod Withdrawal Moderator-Reflector Water Fill On               -        Prevents, Control .Rod Withdrawal Bases The worth of a single fuel pin varies considerably depending upon where the pin is located. Removal of a pin near the center will increase reactivity for under-moderated configurations while removal of a pin on the periphery will reduce reactivity. A maximim worth is specified to provide additional marginito *he limit'of 60 cents&#xfd; excess reactivity in any experiment that removes a fuel pin. Limiting worth to 20 cents also ensures that the operator will not have difficulty controlling power during the normal operation of measuring reactivity changes by pulling control rods to the top stop and measuring reactor period.,
The minimum numnber of four control 'rods is specified to ensure that there is adequate shutdown capability even for the stuck control rod condition.
The insertion time of less .than 1.5 seconds for each control rod from its fully withdrawn position is specified to ensure that the insertion time does not exceed that assumed when analyzing the consequence of the most severe credible accident. Experience shows that rod drop time of less.than 900 milliseconds is typical, therefore 600 milliseconds of the total 1.5 second drop time is allocated to. instrument response.
The auxiliary reactor scram is specified to assure' that there is a secondary mode of shutdown available during reactor.operations. The requirement that negative rieactivity be introduced in less than one minutefollowing activation of the, scram is established to minimize the consequences of any potential power transients.&#xfd; The maximum water height of 10" above the top of the core ensures that the water dump will, insert negative reactivity within one minute of activation, provides: alarge upper reflector to allow consistency between critical position measurementsand&#xfd; experirhents, and prevents instrument tube' flooding: that could disable a safdty -system,channel.
The safety system channels listed'in Table 1 p'rovide a highi degree of redundancy to assure that human or mechanfical failures will not endanger the reator facility or the general public.                                                                    ,,,
The interlock system listed in Table 2 ensures that only authorized personnel can operate the reactor and the proper sequence of operations is performed. It also limits the actions that an operator can take, and assists the operatqr in safely operating the reactor. The minimum flux level has been established at,2 Cps to prevent a source-out.startup and.
provide a positive indication of proper instrument function before any reactor startup.
The annual limit, for integrated power is set at 2kWh to'ensure that the maximum dose in any unrestricted area will not exceed 100 mrem per year and the maximum dose in any restricted.area (not including the reactor room itself,; which should not normally be occupied during operation) will not exceed 5 rem per year.
3.3      Coolant systems - None required 8

Technical Specifications 3.4:     Containment or confinement - None required 3.5'     Ventilation Systems     - None required 3.6      Emergency Power - None required 3.7      Radiation Monitoring Applicability; These specifications apply to the miflimum' radiation monitoring requirements for reactor operations.,:
Ohje tive:              .   ,.
The purpose of these specifications is to ensure that adequate monitoring is available to preclude undetected radiation hazards or uncontrolled release of radioactive material.
Specifications
: 1. The min'imum 'conripleniert'of r~diation m'onitoring'equipifient required to be operating for reactor operation shalt iriclude"
: a. A criticality detector system that m6nitorstl&#xfd;e main fuel storage area and also functions as an area :monito'r' Thissy&#xfd;se m shall haVe'a visible and an audible alarm in the cont-rol room.'
: b. *Anarea gamma. monitoring system that:shall have detectors at least in.the following locations: (1)..control room; (2) reactor room near the fuel vault; (3) reactor~room (high level nionitor),-and; (4) outside~the reactor room window.,"                          ,       ,
: c. The radiation monitors required by 3.3.:1-a and b, may. be temporarily removed-from serviceif replaced. by an equivalent portable unit.
: 2. During normal operation.; acalibratedand operational, portable survey meter capable of measuring ambient radiation exposure will be available.
: 3. During fuel loading or irloading, or, durihg any experiments involving the addifiornior removal of niaferial frn the cbr' (activation foils, etc.) a thin-window GM detector will be available to check for personnel or area contamination.
Bases The contifno'uus- m-onitoring 'of radiationlevlsin tihe reactor room and other stations ensures the waxrning of the existence of anyabnormally high radiation levels. The "
avai abitity bf instrumreiits to' m'easi:rd t' daiiiouht df particulate activity in the' reactoIr room: air,ensures continued compliance-with the requirements'of 10 CFR Part 20.: The availability of required portable monitors provides assurance that personnel will be able to monitor potential radiation fields before an -area'is. entered and during fuel handling.
In all cases, the low power levels encountered in Op~ratio 4of the critical assembly minimizes the probable existence of high radiation levels.
                                                 '9

Technical* Specifications 3.8      'Experiments' Applicability Thesespecifications aipp* to alllexperiments'placed'in the reactor tank.
Olijective The objective of these specifications is to define a set of criteria for experiments to ensure the safety, of the reactor and personnel.:
Specifications
: 1.       No new experiment shallbe performed until a %witte'nprocedure. that has been developed~to permit good understanding of the safety aspects is reviewed and approved by the Nuclear Safety Review Board and'approved bythe Operations Supervisor. Experiments that fall in the general category,. but with minor, deviations from those previously performed, may be approved directly by the
       ,Operations Supervisor.          .
: 2.        No experiment shall be conducted if the associated experimental 'euipment could interfere; with, the control rod functions or with the safety functions, of the nucleair instrumentation.
: 3.        For movable experi ient with an absolute Worth greater than $35,'themaximum reacti,)ity' change fbr withdrawal and insertion ;hall be $.20/sec. Moving parts worth less than $.35 may be oscillated at higher frequencies in the :core.
: 4.        The maximum positive step inseqi0hn of reactivity that can be cau'sed by an experimental *ccident or'expeeriihhtal equipm;erit filure.of Amo vable or unsecured, experiment shalal:not excee'd $.60:.-
: 5.   -      xperiments shall not contain iaterials which can cause a          1io ent chemical reaction. Unencapsulated experiments shall not contain a materiaI that may.
produce significant airborne radioactivity. Encapsulatedd-xperinients may contain materials that can cause a::minor release of airborne radioactivity, subject tothe.
S     *limits in Technical Specifications&#xfd; 3.8.8.
: 6.       'Experiments containing kownexploSives or highly flammable mnatenas"shall not be in sAiued irn the rea'c r .. .              ..    '  "
: 7.        All experiments that corrode easily and are in contact with the moderator shall be encapsulated within corrosion resistant containers.                          '
: 8.        All experiments containin"g radioactive matefial shall be evaluated' forth'iri
        .potential release, of airborneradioactivity and limits shall be 6stablishedfor, the permissible concentration of radioisotopes in the experiments such that a complete release, of all gaseous,, volatile, or particulate constituents to the reactor room air would not exceed the limitations for'exposure of individuals ifi restricted or unrestricted areas.
Bases 10

Technical Specifications The basic experiments to be performed in the reactor programs are described .in the Safety Analysis Report (SAR). The present programs are oriented toward reactor operator training, the instruction of students, and with such research and development as' is permitted under the terms of the facility license. To ensure that all experiments are well planned and evaluated prior to being performed,, detailed written procedures for all new experiments must be reviewed by the NSRB and approved by the Operations Supervisor.
Since the control rods enter the core by gravity and are required by other technical specifications to be operable, no equipment should be allowed to interfere with their functions. To ensure that specified power limits are not exceeded, the nuclear instrumentation must be capable. of accurately monitoring core parameters.
All new reactor experiments are reviewed and approved prior to their performance to ensure that the experimentaitechniques&and procedures.are safe and proper and that the hazards from possible accidents are minimal: A maximum reactivity change is established for the'remote positioning and for oscillation of experimental samples and devices during reactor operations to ensure that the reactor controls are readily capable of controlling the reactor.
* All experimental apparatus placed in the reactor muist be properly. secured. In ,
consideration of potential accidents, the reactivity effect of movable apparatus must be limited to the maximum accidental step reactivity insertion analyzed. This corresponds to a 0.60$ positive step while operating at full power follo,,ed by one failure in the reactor 'safety system.
Restrictions on irradiations of explosives and highly flammable materials are imposed to minimize the possibility of explosion' bf fires'in the vicinity of the reac'to'r.
To minimize the possibility of exposing. facility, personnel or the public to radioactive materials, no experiment will be performed with materials that could result in a violent chemica 'reactiin, prod'u'ce airbir&e activity, or  6 cause a corrosive ~attack on the fuel cladding'or pfimniryr"&codlafit system.         ',
Specification, 8, will ensure that the quantities of radioactive. materials contained in experiments will be so limited that their failure will not result-in exposures to individuals in restricted.or unrestricted areas to exceed the maximum, allowable exposures stated in 10 CPRR20.' The 'estric-ed area miiaximum is defined'in 10'CFR 20.1201 through 10 CFR 20.1204. The unrestricted area maximum is defined in 10 CFR 20.1301 and 10 CFR 20.1302.         - . '                        '                    ",.
3.9      Facility-sPecific Limiting Conditions for Operations All fuel transfers shall be conducted underothe direction!of a licensed senior reactor operator.:                              .,
Operatihg personnel shall 'be familiar with' health physids procedures and monitoring techniques, and shall mofiitor'the ope'taii"nr'with 'ap'propriate radiation instrimentation.
For a completely unknown or untested core, fuel loading shall follow the inverse multiplication approach to criticality and, thereafter, meet Specification 4.2. Should any*.
interruption of the loading occur (more than four days), all fuel elements except the initial 11

Technical Specifications loading step shall be removed from the score in reverse sequence and the operation, repeated.
For a known core, up to a quadrant of fuel pins may be removed fro'mrthe core or a single stationary fuel pin be replaced with another stationary pin only.under the following conditions:
: 1. The net change in reac ftivity has been p 'eViously  deterhiined by measurement or calculation to be negative or less than 0.20$.
: 2.      The reactor is subcritical by at least 1.00$ in reactivity.
: 3.      There is initially, only one vacant.positi~onwithinthe, active fuellatti~ce.
: 4.      The nuclear instrumentatibnii on 'scale and ihe'dump-valve is hot bSpassed.
: 5.      The critical rod bank position is checked after the, oPerationisscomplete....
: 4.      SURVEILLANCE REQUIREMENTS 4.1     Reactor Core Parameters Applicability These specifications apply tothe verification of shutdown reactiyity, reactivity worth of fuel, and reactor power levels that pertai.nto r~eactor control.
Ohjective The purpose of these,,specifications is, to ensure, that the analytical bases are~andxremain valid and that the reactor is safely operated..
Specifications The following parameters shall be determined during the initial testing of an unknown or previously untested core configuration:
: a.      excess reactivity;
: b.      worth of most reactive fuel.pin;    ,,.,
: c.      reactor power measurement;'
: d.      shutdownreacti'vity.
Bases:-
Measurements of the above paramreteis are made *hen anew reactor configuration is assembled.. Wheneverthe core.configuration is altered to. result in an unknown.-dif*
untested configuration, the core parameters, are ,evaluated to ensure that theyvaire within the limits of these specifications and the values analyZedin the SAR. During this test period of the reactor, measurements are performed using the approved experimental procedures.
12

Technical Specifications The excess reactivity measurement ismade to~verify that thisconfiguration is not subject to a reactivity addition accident more severe that that analyzed and described in the Safety Analysis Report, Section 13.2.
This same accidentassumes a scram signal at a maximum power level of 100 watts indicated so it is necessary to measure reactor power and make any necessary adjustments to the instrumentation that indicates reactor power. The scram signals are based in detector current while the vfsual display is in watts. The high current scram must be verified to not exceed an indicated 100 watts.
Lastly, the accident analysis assumes the reactor is shutdown by at least $1.00 reactivity after the high ctrrfent scram occurs.,Shutdown reactivity is also measured to ensure the reactor meets, the definition of, shutdown when all .control rods are on the bottom.
4.2      Reactor Control and Safety Systems Applicability These specifications apply to the surveillance of the safety and control apparatus and instrumentation of the facility.
Ohjective The purpose of these specifications is to ensure that the safety 'and control' equipment is operable and will function as required in Specifidatioh'n33.2.
Specifications
: 1.       The total'control rod!drop'timne; -including instiu-mre nt response time shall be' measured semiannually to verify that thefrequirenients of Specification 3.2, Item 3, are met.
: 2.       The moderator-reflector water dump time shall be measured semiannually to verify that the requirement of Speificatioh 3.2, Item, is met.
: 3.       All safety system channels shall be calibrated annually.
: 4.       A channel test of the safety system channels (intermediate, and power range instruments) and a visual inspection of the reactor shall be performed daily prior to reactor startup. The interlock system shall ber checked daily prior to reactor startup to satisfy rod drive permit. These systems shall be rechecked following a shutdown in excess of 8 hours.
: 5.       The moderator-reflector water height shall be checked visually before reactor startup to verify that the requirements of Specification 3.2, Item 5, are met.
: 6.       These tests may be waived when the instrument, component, or system is not trequired to be.operable, but the instrument, component or system shall be tested prior to being declared operable.
13

Technical Specifications Bases Past performance of control rods and control rod drives and the moderator-reflector water fill and dump valve system have demonstrated that testing semiannually is adequate to ensure compliance with- Specification"3.2, Items .3and 14.
Visual inspection of the reactor components, including the control rdds', prior' to each day's operation, is to ensure that the components have not been damaged and that.the core is in the proper condition. Redundant safety channels are provided by having three independent channels provide higfh currehits'&#xfd;rams if nacessary and by requiring all three channels be operable. The analysis of the most severe accident sh6ws':nd fuel damage, even if one channel fails. Random failures should not jeopardize the ability, of the overall system to perform its required functions. The interlock system for the reactor is designed so that its failure places the systemin a safe or non-operati.ng condition,. However,-to ensure that failures in.the safety, channels: and interlock system are detected .as soon as possible, frequent surveillance is desirable and thus specified. Allof the above",
procedures are enumerated in the daily startup checklist.
Past experience has indicated that, in conjunction with the daily check, calibration of the' safety channels annually ensures the.proper accuracy is maintained.
4.3      Coolant Systems                 to Applicability                                        .
These specifications apply to moderator in the storage tank or reactor tank.
Oijective
                                ' ' I ,. -'-     *                         ......   .. '
The purpose of these specifications is to ensure the continued validity' of radiation protection standards in the facility.                            .
Specification                                           .
Analyze moderator for'radioactivity prio.r to, discharge.to the environment.,
Bases..                              "       ' .   ,
Experience has demonstrated thai the moderator does not accumulate radioactive material due to the low operating neutron fluence.. Therefore, periodic monitoring is not~necessary.
Verification is necessary, how'ver,.prior to d scharge to th-e-*        * : ..
thuienvironment       . ,i" -..
4.4      Containment or Confinement - None required 4.5. Ventilation Systems - None required; 4.6 -Emergency         Power - None required 14

Technical Specifications 4.7     Radiation Monitoring Applicability These specifications apply to the.surveillance of the area radiation monitoring equipment and all portable radiation monitoring instruments.
jective    .-
The purpose of these specifications is to ensure the continued validity of radiation protection standardsinn the facility.,
Specification.
The criticality detector system,&#xfd; and area gamma monitorsshall be tested with a radiation source at least~mdnthly and daily if the reactor is -operatedand calibrated semiannually.
Portable instruments: shall be. calibrated annually."
Bases Experience has demonstrated that calibration of the criticality, detectors and gamma monitors semiannually is adequate to ensure that significant deterioration in accuracy does not occur. Furthermore, the operability of these radiation monitors is included in the daily pre-startup checklist. If the reactor is not operated for more than a month, the instruments are required to be checked to ensure operability. Portable instruments are calibrated at the manufacturer recommended frequency.
4.8     Experiments - None required 4.9     Facility-specific Surveillance Requirements - None required
: 5.      DESIGN FEATURES 5.1     Site and Facility Description The facility is located-on a site situated on the 'Suth'bank of,the:Mohawk River in the City of Schenectady. An inner fence of greater than 30 feet radius defines the restricted area. An outer fence and riverbank of greater than 50 feet radius defines the exclusion area.
The facility'is housed n'in the reactor building. The Security of the facility is maintained by the use of two fen ces; IoIne at the site bundary and'ihe oth&r dfiningthe restricted area around the reactor building itself.
The reactor room is a 12-inch reinforced concrete enclosure with approximate floor dimensions of 40x30 feet. The height from: t*hel ground fibor.lto thec~iling shall, be about 30 feet. The roof is a steel deck covered by 2 inches of lightweight concrete, five plies of felt and asphalt, with a gravel surface. Access to the reactor room is through a sliding fireproof steel door that also contains a smaller personnel door. Near the center of the room is a pit 14.5 x 19.5 feet wide and 12 feet deep with a floor of 18-inch concrete.
15

Technical Specifications This pArt- contains the 3500,gallon water storage tank and. other piping and. auxiliary equipment..
5.2      Reactor Coolant System The reactor core is installed m a stainless steel reactor tank that has a capacity of approximately 2000 galons' of water. :The tank nbomin'al dimension's are 7 feet in diameter and 7 feet high: The tank is u'&#xfd;ppofted at floor level above the reactor room by 8-inch steel I-beams. There are no side penetrations in the reactortank.
The reactor tank is connected to the water storage tankvia'a &#xfd;ix-inch quick dump line.
Therefore, it is required that the storage tank be vented to the atmosphere such that its freeboard volume can always contain all water in the primary system. The water handling system allows remote filling and emptying of the reactor tank. It provides for a water dump by means of a fail safe butterfly-type gate- valve when a reactor scram is initiated.
The filling system shall be controlle1 by'.the operator, who must satisfy the sequential' interlock system before adding water to the tank. A pump is provided to add the moderator-reflector water from the.storage dump tank into the reactor tank. A fast fill rate of about 50 gpm is provided. A nominal six-inch valve'is installed in the dumpline and has the capability. of emptying the reactor tank on demand of the operator or when a reactor scram is initiated, unless bypassed with the approval .of the licensed senior operator on duty. A valve is installed in the bottom drain line of the reactor tank to provide for completely emptying the reactor tank.
5.3      Reactor Core and Fuel The ireac'tor core shallfonsist of u;aniu      fudi*in the form of 4.81 Weight percent or less enriched U0 2 pellets in metal cladding, arranged in roughilya' cylinidrical fashion with four control rods placed syinmet iCally-ab6ut the core periphery. 'The 'total core configuration and the arrangement of individual fuel pins, including an ylexperiment, shall comply wfith'the re               ofItiese'Technical Specifications fodnd in Sections fuir~ments 3.1 and 3.2 of this license. The core shall 6on'sistof all SPERTY F-1) fuel"described in 5.4.3.                                    '
The fuel pins are supported, and positioned on a fuel pin supporf plate, drilled with holes to accept tips on the end'of each pin. The support plate rests on a carrier plate, which forms the base of a three-tiered overall core support structure. An uppei'fuel'liaftice plate rests, on the top plate, and' both are..drilled.through with holes withthe prescribed,;
arrangement to. accommodate the upper ends of the~fuel pins. Thelower fuel pin,support plate, a middle plate, and the upper &#xfd;fuelpin,.lattice platea.re secured with tie rods ,and bolts. The entire core structure is supported vertically and anchored by four.,posts set in the floor of the reactor tank.
Core fuel, pins to&#xfd;ebc utilized are 4.81: weight percent enriched, SPERT (F-i,) fuel rods.
Each fuel rod. is~made.up of sintered U0 2 pellets ,encased ina stainless steel tube, capped on both ends with a stainless steel cap and held in placewith a chromium nickel spring.
Gas gaps to accommodate fuel expansion are also provided at both the upper end and around the fuel pellets. Figure 4.5 of the SAR depicts a single fuel pin and its pertinent' dimensions.,NUREG-1281 describes thesefuel pins in additional' detaiL-16

Technical Specifications Four controlrod;assemblies are installed; spaced 90 degrees apart at.the:core periphery.
Each rod consists of a 6.99-cm square stainless steel tube, which passes through the. core and rests on a hydraulic buffer on the bottom carrier plate of the support structure.
Housed in each of these. "baskets" are two neutron-absorber sections, one positioned above the otheras depicted h Figure 4.6 of the SAR. The combination of the four rods must meetfthe values given in Table' 13.2 of th& SAR, with regard to reactivity with one stuck rod and shutdown reaci'ivty.
5.4      Fissionable Material Storage When not` in Use, the SPERT(F-S           1) fuelshall be stored withih the storage vault located in the reactor room.' The vault shail be 'closed by a locked door and shall be provided with' a criticality monitor near the vault door: Thefuehlsghall be stored in cadmiium clad steel tubes with'no mor6`fthan 1 kg'fuel per tube miounted on a s~teel wall rhck. A storage tube in the''st6rage vault-canhot 6ontdiin more thani 15 SPERT(F-1).frel pins at anytime. The center-to-center spacing of the storage tubes, together with the cadmium clad'steel tubes, ensures thatthe infinite multiplication factor is less than' 0.9 When the vault is fully' flooded with water.
: 6.       ADMINISTRATIVE CONTROLS 6.1      Organization 6.1.1    Structure The ,organization for the management and operationof the reactor facility shall include the structure indicated in Figure 6.1L Level 1:          The Dean, School of Engineenng&#xfd;,appoints.the Chair, Nuclear Safety Review Board.
Level 2:         ;.The Faclity Directp reports to the:Chair.,, Mechanical, Aerospace and Nuclear Engineering for administrative purposes...
Level 3:          The Operations Supervisor reports to the Facility Director.
Level 4:          Licensed operators and senior operators are the operating staff and report to the Facility Director for' adhinistrative pu'+poses.
6.1.2    Responsibility,,
The Dean, School of Engineering', is responsible for the facility license and appoints the Chair;,Nuclear Safety Review Board.,The Facility Director is resp6nsible for facility' -
adminitsration: aind safety., The Operations:Su pervisor is responsible for the day-td-day safety. and-operation' of the facility.,.
The RPI Radiation Safety Officer (RSO) who is organizationally independent of the reactor*operations grdou shall provide'advice as required bynthe Facility Director and the Operations-Supervisor in matters ,concerning radiological safety. The RSO also has interdiction responsibility and'authority.
6.1.3    Staffing
                .19 (a)      The minimal staffing when the reactor is not. shutdown as described in these! -
specifications shall be:
17

Technical Specifications
: 1)       An operatoror senior operator licensed pursuant. to0.O CFR 55 present at the controls..                          .
: 2)        One other person in the control room certified by the Reactor Supervisor as qualified to activate manual scram and initiate emergency procedures.
This person is not requiredif an operator and a senior operator are in the control room.
: 3)        A licensed senior operator shallibe present or readily available on call. The identitylof and method for. rapidly contactingthe licensed senior operator on duty shall be known to the operator.       -'":, !., . .- t :,
(b)     Theminimal staffing whenthe reactoris::shutdown,,but-not in safe shutdown is a senior reactor operator in the control room and a second senior;reactor operator present or readily available: on call..
(c)     A list of reactor facility personnel by, name and telephone. number shall be readily available in the control room for use by the operator. The list must include:
1).; ' :Management personnel.-
: 2)        Radiation safety personnel.
: 3) &#xfd;i Other operations personnel.
(d)     Events irequiring the direction,of the Operations!Supervisor: -
: 1)       All fuel or control rod relocations within the reactor core unless the activity,,iS part of an approved experiment,
: 2)        Recoveiy from uriplanfned or unscheduled'shutdown'.
6.1.4   Selection and Training of Personnel, The..selection, training and reqiialification.' of operations personnel shall meet or exceed the requirements 0f AmericanNational Standard for Selection and Training of Personnel for Research Reactors, ANSI/ANS-15.4-1988, Sections 4-6.
Additionally, the minimum requirements for the Operations Supervisor are atleast four years of reactoi' opefating experienceanfd possession of a'Senior Operator License for the RPI Critical Facility. Years spent in lbaccataureatolor'graduate study, may'be substituted, for operating experience on, a one-fot-onhe basis' up to a mfaximum of tw 6 years.'
6.2     Review and Audit                    .
A Nuclear Safety Review Board (NSRB) shall reyiew and audit reactor operations and advise the Fa'ility Director in matters relating to. the health and safety of tihe',public and, the safety of facility operations.            .-
6.2.1             Composition and Qualifications TheNSRB shall be appointed by the Dean Schoolof Engineering in' accordance. with the NSRB Charter.           '                                    '        .             '
6.2.2             Charter and Rules The NSRB Charter shall describe the composition of the board. The Review Board shall function under the-following rules:
(a)     TheBo3ard shall meet atleast semiannually.
(b)     The quorum shall consist of not less than a majority of the full Board ahid shall include the Chairman or his designated alternate.
18

Technical Specifications (c)   I Minutes of each Board meeting&shall be distributed to the Dean, NSRB members, and such others as the Chairman may designate.
6.2.3             RevieW Function' The following items shall be reviewed and approved by the NSRB-before implementation:
(a); Proposed- experiments and tests utilizing the.reactor facility that are significantly different from tests and experiments previouslyperformed at the facility.
(b)      Reportable occurrences.
(c) ,     'Proposedchangestothe'TechnicalSpecifications and-proposed amendments to facility license,--    .              :,
(d)      Operating, Emergency and Surveillance procedures.;
6.214             AudifiFunc"ti6n.
(a)      The audit function shall include selective (but comprehensive). examination of operating records, logs, and other documents.,, Where necessary, discussions with cognizant personnel shall take place. In no case shall the individual immediately responsible for the area auditin the area. The following areas shall be audited at least annually'.
(b)      Reactor operations and reactor operational records, for: compliance with internal rules, regulations, procedures, and with licen.sedproyisions; (c)      Existing operating procedures for adequacy and to ensure that they achieve their intended purpose in light of any changes's'in'e thei-i iinple'ientati0n;.
(d).    -Plant equipment performance. with particular,attention~to operating anomalies,
        -abnormal occurrences, and the~steps'taken to'identifyvand, correct their use,.
6.3      Radiation Safety The Radiation and Nuclear, Safety Qommittee and the Radiation Safety,. Officer shall be responsible-forythe implemeIntat!0n of the Raddition .SafetyProgram for the RCF. The,'
primary purpose ofthe,program is to, assure radiological-safety for all University personnel and the surrounding community.
6.3.1    AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) PROGRAM Control 6f ionizing radiation exprosure is based on the assumpption that any exposure rnvolve's some risk. However, occupational exposure within accepted limitsrepresents a very small risk compared to the other risks voluntarily encountered in other work environments.                                 ..-      .       _
The policy of Rensselaer Polytechnic Institute is to maintain occupational exposures. of individuals to be well within allowable limits as are defined in the appropriate,.
regulations. The individual and collective dose to workers in maintained; as low as reasonably achievable (ALARA).
AiARA is a part'of the normal Work process whe're people are working*with ionizing radiation. Management at all levels, as well as each individual worker, must take an active role in minimizing this radiation exposure.
19

Technical Specifications Exposures at the facilityiare routinelyreviewed byitheRadiation Safety Officer and Radiation and Nuclear Safety Coinmittee to ensure thatproper radiation safety procedures are in place and ALARA is maintained.
6.4      Procedures Written procedures shall be prepared, reviewed and approved -prior to initiating any of the activities listed in this section.. The-procedures, including applicable check lists, shall be reviewed by the NSRB and followed, for the following operations:,
: 1)       Startup, operation and shut down of the reactor..
: 2)       Installation and removal of fuel pins!, control.rods,: experiments, and experimental facilities.                                            and
: 3)       Corrective actions to be taken to correct specific andforeseen malfunctions such as for power failures, reactor, scrams, radiation, emergency, responses to alarms, moderator leaks and abnormalreactivity changes..
4),      Periodic surveillance of reactor instrumentation and safety systems, area monitors, and continuous air monitors.
: 5)       Implementation of the facility security plan.       '.   "'
: 6)       Implementation of facility emergency plan in accordance with 10 CFR 50, Appendix E.
: 7)       Maintenance procedures that could have an effect onf reactor safety.
Substantive changes to the aboVe procedures shall b6 made only with the piriorajpproval of the NSRB; Temporary changes to the procedures that do not change their.original intent may be made 'with the app'roval of the Operatibns Supervisor. All such temporary changes to the proceduies sha'lFbe documente'd nd subs~queiitly reviewed by the Nuclear Safety Review Board.
6.5      Experiment Review: and Approval ,           .    *.
: 1)       All new experiments or classes of experiments that might irivolvean unreviewed safety question shall be reviewed by the Nuclear Safety Review Board.' NSRB,,
approval shall ensure that compliance with the requirements of the license technical specifications, and 10' CFR50.59 shall be documented.. A: licensee shall.
obtain a license amendment pursuant tSec. 50 90 prior toimplementing a, proposed change, test, or experiment, if.the change, test; or experiment would:.
(a) Result in more than a minimal increase in the frequency of occurrencze of an"'
accident previously evaluated in the final safety analysis report;
        .(b) Result.in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important. to, safety previously evaluated in the final safety analysis report; (c)" Result in more' than a minimal increase in the consequences of an acident ireviously evaluated inthefinal'safety analysisreport; (d) Result in more than a minimal increase inthe consequences of a malfunction of aan SSC important to safety previously evaluated in the final safety analysis report; (e) Create a possibility for an accident of a different and potentially more severe than any previously evaluated in the final safety analysis report; 20

Technical Specifications (f) Create a possibility.for a malfunction of an SSC important-to safety with a different result than any previously evaluated in the final safety analysis.
report; (g) Result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered; or (h) Result in a departure from a method. of evaluation described in the SAR used in establishing the design bases or in the safety anaryses.
: 2)        Substantive changes' topreviousl5y approved'experiments shall be made only after:
review and approval in w.riting by NSRB. 'Minor changes, that do not significantly alter the experiment may be'approved by the Operations Supervisor.
: 3)      "Approvedexperiments: shall be cafried&6Ut in accordance with established approved procedures.
4)',      Prior torr&view, an experiment planAor proposal shall be prepared describing the eXperimeiYt,.including any safety considerations. .
: 5)        Review comments of the NSRB s'etting, forth ahy conditions and/or limitations shallibe documented in committee minutes and submitted to thel Facility Director.
6.6       Required Actions in the Event of a Reportable Occurrence
: 1)       The reactor shall be shut down. Operations shall not be resumned u'nless authorized by the Chair, NSRB.
: 2)        Occurrence shall be reported to the Facility Director or designated alternate, the NSRB and to the Nuclear Regulatory C6mmission hot later thanthe followinig.
working day by ielephone and confird in wrtingio licensing authorities, to be
         'folloWed by a written report that i.describ:es thl cir6uhnsiances of the event within 14 days of the event:.
: 3)       All such conditions, including action taken to prevent or reduce the probability of a recurrence, shall be reviewed by the NSRB. The NSRB shall concur with corrective, actions ...
6.7       Reports                                .                       .
In addition to: the requirements of-applicable regulations, and in no way substituting therefore, all written reports shallbe-sent'to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy, to the Region I Administrator.
6.7.1 Operating Reports A written report' overing the6previous year shall b* submitted by March 1 of each year.
It shall        diudu'the following:
(a)       Operations Summary. A sumniiaryo6f operating experience occurring during the r.eprting period that relates to th* safe: operation of the facility, including:
1).      Changes in facilityd sign-          .
2)'      Performance characteistics (e.g., equipment.and fuel performance);
3)-:     Changes in operating procedures ttiit' relate to the safety of facility operations; 21

Technical Specifications
: 4)       Results* of surveillance tests and inspections required by these Technical Specifications;
: 5)       A brief summary of those changes, tests, and experiments that require authorization from the Commission pursuant to, 0 CFR,50.59, and;,
: 6)       Changes in the plant operating staff serving in the follQwing positions:
a)       Facility Director; b)       Operations Supervisor; c)       RSO; d)       Nuclear SafetyReview Board Members.'-
(b)    Power Generation. A tabulation of the in'-tegrated therm'al p~wer 'ding the reporting period.     '         .              '  .                         ,
(c)     Shutdowns. A listing of unscheduled shutdowns that,have occurred during the reporting period, tabulated according to cause, and a brief description of the preventive action taken;to prevent ,recurrence..      '.-                            ,
(d)    Maintenance. A tabulation of corrective maintenance (including major preventative maintenance) performed during the reporting period on 'safety related systems and components.                    . ...
(e)     Changes, Tests and Experiments.;, A brief description and, a summary- of the safety evaluation for all changes, tests, and experiments that were carried out without prior Commission approval pursuant to the:-requirements of,10 CFR Part 50.59.
(f)    A summary of the nature, amount and maximum concentrations of radioactive effluents released or discharged to the environis beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.-
(g)    Radioactive Monitoring.. A, summaryof~the TLD doseratestaken at the exclusion area boundary and the site boundary during the reporting period.
(h)    ,OccupationalPersonnel Radiation Exposure. A summary. of radiation exposures greater than 25% of the values allowed by 10 CFR. 20 received during the reporting period by facility personnel (faculty, students or experimenters) and visitors.
6.7.2   Special Reports                                          .
(a)    Reportable Operational Occurrence Reports. Notification: shall be made within 24 hours by'telephone in accordance with 10CFR50.36(c)(7) followed bya written, report in accordance with LOCFR50.36(c)(5) within 10 days in the event of a reportable operational occurrence as defined in Section 1.3. The written report on these reportable operational occurrences, and to the extent possible, the preliminary telephone and e-mail notification shall: (1) describe, analyze, and evaluate safety implications; (2) outline the measures taken to ensure. that the cause of the condition is determined; (3) indicate the corrective action (including any changes made to the procedures and to the quality assurance program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems; and (4) evaluate the safety implications of the incident in light of the cumulative experience obtained from the record of previous failures and malfunctions of similar systems and components.
(b)    Unusual events. A written report in accordance with 10CFR50.36(c)(5) shall be submitted as specified in 10CFR 50.4 within 30 days in the event of discovery of any substantial errors in the transient or accident analyses or in the methods used 22

Technical Specifications
: for suchanaly'ses, as described in*the Safety -Analysis Report or in, the bases for the Technical Specifications.
(c)       Key changes'in'org'anization. A written, reportin accordance with 10CFRM50.36(t')(5) submitted*as specified in 10CFR 504 shall be provided for any changeo'inLevel 1 or'Level 2' personnel.,
6.8       Operating Records 6.8.1     The following records. and logs shall be maintained at the Facility or at Rensselaer Polyt~chnic.I!nstitutp for at east five ypags.
(a)       Normal facility operation (except retain checklists for one year) and principal
        !,maintenance operations;-
(b)       reportableoccurrences; (c)       tests, checks, and measurements documenting compliance with surveillance requirements; *                             .
(d)       experiments performed with the reactor; (e)       fuel shipments, inventories, and receipts; (f)   :   reactor facility radiation and contamination surveys; (g)       approved changes to operating procedures,;
(h)       records of NSRB.meetings and audits:
6.8.2' Records to be retained for at least one certification cycle' Records of training or retraining of certified operations, personnel shall be&#xfd; maintained at all times the individual is employed or until the certification is renewed.                                  '                            -
6.8.3     The following records and logs, shallibe miaintainred at the acility or' at Rensselaer for the ife of the Facility.
(a)       gaseous and liquid radioactive releases from the facility; (b)       TLD environmental monitoring systems; (c)       radiation exposures for all RPI Critical Facility personnel (studenfits and experimenters) and visitors;                    '                         ,                 '
(d)     . the present-as-builit facilitydrawings and new :updated or corrected versions.
            .i -:  ; 3 * " , ', ' , , ;.         ".    :        : . ". ' * ''         a':: - * . i '! *   ' 'a.". , . '
G                                                               j
                  ' t.; . .' .- &#xa2; *-,
                                  *   -; *. , ,'a,      * ,. ; . ,   ..  . -" "a      i - : z * , "     .  '  , .       "
23

Technical Specifications Vice Presicdenlt, Human Resources;&#xfd;'
Direclor, Enviio'nment, Health & Safety Chair, Mechanical, Aeronautical &
Nuclear Engineering g
CDirector, Critical Experiments Facility' Figure 6.1:- RCF Management Qrganization "t, I, *   '" .* '
24

Explanation of Changes to Technical Specifications The changes explained herein are relative to the current Technical Specification, valid for Amendment 11 to the license. The changes then are differences in the proposed Technical Specification.                          ,     .
References used in the explanations are:.           .
: 1. NRC Request For Additiofial Inf6rmatio6, March'21, 2008
: 2. Technical Specificatons (current version), datedSeptember 2004 Note that Ref. 1 uses tet'an'd 6rganization of a third version of the Technical Specifications, specificallfy the version submitted in 2002 with aniapplication for relicensing. Except forresponseslo'Ref. 1, this version of the Technical Specifications is no longer in use.
The changes relative to Ref. 2 are:
: 1.       Section 1.1 added.
: 2.       Section 1.2 added to show which version of ANS-15.1 was used.
: 3.       Section 1.3'Definitions were imported fromANS-15.1-2007 and are not numbered as in Ref. 2. This is numbered Section L1,0 in:Ref. 2. Specific definition changes are presented below. Definitions added as p1art 6fANS-15.1-2007 are not discussed and no justification is considered to b6 require'd. Note that references to specific definitions in later sections of the Technical Specification have been revised to accommodate the revised format,andnumbering.
: 4.       Definitions added for certificate or charter, certified, Class A reactor operator, Class B reactor operator and channel.
: 5.       Definition of channel calibration changed to agree with ANS-15.1-2007
: 6.       Definition of channel check changed to agree with ANS-15.1-2007.
: 7.       Definition of channel test changed to agree with ANS-15.1-2007.
: 8.       Definition of control rod assembly (Section 1.0.D) changed to control rod and reference to materials not in use was removed from the definition. There is no plan to use control-rods~othertfian those in the definition.
: 9.       Definition added for core configuration.
: 10.      Definition of excess reactivity changed to agree with ANS-15.1-2007.
: 11.      Definition of experiment changed to agree with ANS-15.1-2007.
12,      Added definitions for facility-specific definitions, known core, license, licensed, and licensee.
: 13.      Removed definition of measuring channel (Section 1.0.G) as redundant with the definition of channel.
: 14.      Definition of measured value changed to agree with ANS-15.1-2007.
: 15.      Definition of movable experiment changed to. agree with ANS-15.1-2007.
: 16.      Added definition of owner or operator.

17            -Definition of operable:changed to'agree With ANS- 15.1-2007,.
: 18.              Definition. of'operating dhanged~ o agree With ANS-15.1-2007.
: 19.              Added. definition ofpermit...-
: 20.              Added definitions of protective:action; reactor operating and reactor operator.
Reference to xenon in the ANS -15.1-2007, definition of reactor operating was removed since the RPIreactor does&#xfd; not..generate&#xfd; detectable xenon:,,
21.'             Definition of reactor safety systems changed to aigree with ANS:"15.1-2007.
: 22.              Definition of reactor secured. changed to agree more closely. with .ANS- 15.1-2007 and reflect facility-specific conditions.
: 23.              Definition of reactor shutdown changed toragree with ANS-15.1-72007 and to reflect facility-specific conditions..
: 24.            'Added definiti'n-of reactivity worth of 'anexperimdnt&#xfd;
: 25.              The definition for readily available on call was revised. to 60 minutes travel time or 25 miles. The previous definition' was more restrictive, 30 minutes'and 15 miles: The justificati'on for this change was provided with the response to the Request for Additional Information', item 11:6 and-is discussed in the Safety Analysis Report, Chapter 12.
: 26.              Added definition of reference"coie'condition""
: 27.              Added definitions of research reactor and research reactor facility..
: 28.             'Added definition of responsible authority.
29'.             Definition of reportable occurrence revised to blend together the definition in ANS-15.1-2007 and the definition in Ref. 2.
: 30.              Modified definition of safety channel .tolbe consistentwith replacement of the term measuring channel with the term channel.
: 31.              Added definition of scram time..:-
: 32.              Definition of secured &xpririment changed to agre6e with ANS -15j1-2007.
: 33.              Added definition of senior reactor operator.
: 34.          . Added      definition of shall, should. and; may..
: 35.          '   Added definition Of shutd6wn'margin and'supervis6r, rfeactor'operafor.'
: 36.              Changed definition of surveillance frequency to agree* with and to reference ANS-15.1-2007. The sufveillance intdrva[i that are not used at the RPI ieactor were removed from the list."            "     ' :   ' '   "
: 37.              Definition of true value changed 'tc'agree with ANS-15.1-2007.     '  .
: 38.              Definition of unknown core added.
: 39.              Added definition of unscheduled shutdown.               '
     'C. . '       ? ' *.          ,"            "
: 40. Section 2.1 revised to state that no safety limit isrequired., This is consistent with ANS-15.1-2007 for reactors without engineered safety systems. The text for Applicability,.Objective and Specification was deleted. A basis for not requiring a safety limit replaces the text explaining the basis for the current safety limit.-
41., The limiting safetysystem setting for-reactorpower in Section2.2 has been reduced..to 100;watts, a more conservative setting than the 135 watts in the previous Technical Specification. Instrumentation changes make this possible since .high current: setpoints arebased on specific current values rather than 90%
of instrument full scale ranges. The value of 100 watts was chosen to agree with the license -limit.The explanation Of bases was&changed for consistency with the reduced safety system setting.
: 42. Thelimiting safety system setting of 2.0 counts/sec was, deleted. This fnstrument setting is unrelated to safety and is retained as an interlock shown later in the Technical Specifications.. The associated discussion paragraph in Bases was also removed.
: 43. Sections 3.1 and 3.2 were renumbered as 3.2 and 3.1 respectivelyto agree with the organization of ANS- 15.1-2007 and the constraint on excess reactivity moved to the new 3.1 from paragraph 3.1.1 of Ref. 2, Text for Applicability, Objective, and Bases of this specification was added.            .
: 44. Paragraph 3.1.3, Ref. 2 was deleted as unnecessary. The analyzed accident is independent of control rod reactivity rate. Further, responses to Ref. l show that the rate.limited by Ref. 2 is hot achievable. Review'wing chahges'to core configuration include a consideration for accidents more severe than that analyzed and this would include reactivity additions by control rods. Note that deletion'of this item renumbered 4ilfollowingitems.
: 45. Control rod drop time has been increased from that specified in paragraph 3.1.4 of Ref. 21The most severe accident assumes a L5 second delay to insert the control rods. The current specification is shorter than required and ma'y not accommodate instrument response. Further, -there is no part of the accident analysis. that relates to a magnet release time so this specification is eliminated.
Associated discussion in Ref. 2, section 3.1 Bas es was revised to agree with the revised drop time limitation.                   "
: 46. Instrument ranges for Linear Power and LogIN power on Ref: 24 s'ection 3.1, Table 1 were changed:to show the ranges ofthe, current.equipment. A: range for Log N Period was added to this Table.
: 47.                       Table for Building Power changed to Reactor Scram. This better defines the action that occurs if BuildingPower fails.
: 48. Footnotes (a) and (b) have been deleted from Ref.. 2, section 3.1, Table 1, as unnecessary and the remaining two footnotes, currently (c) and (d), renumbered as (a) and (b) respectively. Older instrumentation had bypass ca pability. The current instrumentation does not-..
: 49. Footnote (a) has been deleted from Ref. 2, section 3. 1, Table.2, as unnecessary.
Older instrumentation had bypass capability. The current instrumentation does

not. In addition "less than" signs (<) were added to the specification for Reactor Period value and Neutron Flux. value: for completeness. An explanation for minimum. neutron flux was added to section 3.1 Bases.
: 50. The interlock for Failure of Line Voltage to Recorders in section 3.1, Table 2, was reworded and a minimum value established as proposedby Ref. 1, 3.2.9(A).
The value of 100 volts is based on the equipment operating manual.
: 51. The interlock -forReactor Tank Water Level section 3:1 -Table' 2; was ,deleted.
This is not an automatic interlock, but an administrative control established by operating procedures. Explanation in Bases for moderator water-level has been revised.
: 52. A new paragraph was added to the start of Ref. 2, section 3.1 Bases to' discuss the limitation on fuel pin worth. Thisparagraph was placed firs't to agree with the order the constraints are listed.
531     The specification for integrated thermal power was lowered to 2 kilowatt-hours.
This change is a. result qf evaluating radiation survey values inside and outside of the reactor building. See the 'discqussion in.Ref. 1, 4.8. Whilefacility eiivironment monitors show less 'than minimum detectable6 radiation levels inside and' outside, this is a result of operating power lev'els and scheduile. The reduced integrated power value will limit the maximum posSible dose in restricted areas to less than 5 rern per year, and inuirestricted 'areas to less than tOO millirem iier year if the facility operates at licensed power for the total 2 kilowatt-hours.
: 54. "   The last four paragraphs of Ref. 2, sectioni 3.1 Bases' were removed as!
unnecessary since the specification discussed in, those'paragraphs is; discussed elsewher* (excess reactivity, pin worth and reactor power) or deleted (control rod
: 55. , :;Material in section: 3.2 of;Ref> 2.haS:beeii deleted and replaced by the new Section 3.1. The reactor parameters of Ref,. 2 are notparameters used in the accident analysis and need not be stated as limits or constraints. The accident analysis makes no assumption abput temperature or void coefficients, of reactivity or about initial temperature, other than to use cross-sections based on 20C. Nor does the progress of the accide'ntgene*ate significant temperature changes or voids. The new core parameter of'excess reactivity is the necessary parameter to define. theaccident magnitude, along with the~anticipated response from the reactor safety. system.,          '
: 56. Ref. 2 sections' 3.3'and31.4'are renumbered 3.7 and section 3.4,and:3.8:'
respectively.'New sections;3:3-'3.3G6 have been added. This format' chaihge is in accordance with ANS-15.1-2007..           '          ''
: 57. Paragraph 3.3.1c in Ref. 2 has been deleted as unnecessary. As explained in the response to Ref. 1 7.6the ifistrumient is not cdnsidered to be necessary for adequate radiological protection at the RPI reactorfacility. 'The following paragraph was renumbered.
: 58. Requirements for portable radiological instruments have been'added'as paragraphs'3.7.2 ahd 3.7.3 andtext added 'to theBases to explain 'the purpose.

     - This additionwas proposedbyRef. 1, 3.3.2' .Iin Ref. 2 this is 3.3.2 and is less specific as to what instruments arei required..
: 59.      Specification 5 from Ref. 2, section 3.4 hais added text for clarity. See response to
: 60. Specifications 8 and from Ref. 2, section '3.4 c6mbined into a single se ifcton.:se:Response to Re&#xa3; 1              '**
specificat                             f. 1, 3.4.8(A). The associated text in Bases (last paragraph) has' been revised to~reference a single specification and provide correct references to 'iOCFR20.
: 61. New section 3.9 has be'en' added and material from section 5.6 of Ref. 2 was moved to the new section as proposed by Ref. 1, 5.6(A).
: 62. Secti~n.i 4.1 and-4.2 of were renumbered as 4:2 and 4.1 respectivelyand the titles Srevised to agreewith the foriat of ANS-15.1-20071
: 63. The Applicability and Specifications of section 4.2 of the Ref. 2 were changed to measure those parameters of an unknown core tiat are important to determine if the unkn&6wn core meets' the requirements' of secti0n 3 and to calibrate the safety cchannel instruments. Core configuration changes can perturb the neutron flux at the ion chambers. Unless a power calibration is performed, the litmiting safety system setting of 100 watts may not be met.
: 64.      Ref. 2, Section 4.2 Objective revisedto refer to Section 3.2 vice Specification 3:1.
: 65.      Ref. 2, section 4.2, Specifications .l,2and, 5, revised to reference the correct section and specification.
66.-     Ref'. 2, section 4.2 , specification 3, ie'vised to-requirea annual calibration of only the safety system channels. Calibration of all instrumentation, as is now required, is unnecessary since only*the safety system channels provide reactor protection.
: 67.      'Ref.2; section 4:2, Bases, first paragraph, revised to 'refer to the correct section and specification as proposed in Rf. 1, 4.2(Bi. The secon;d paragraph Was rewittien'iforfclarity'as discussed in response to"Ref. 1, 4.2(C).
: 68. New Section 4.3 through 4.6 added and Section 4.3 renumbered as 4.7 to agree with ANS-15.1-2007 format.,
: 69.      Ref. 2, section 4.3' Specification re-vised to remove the mobile particulate gamma monitor as discussed in the response to Ref. 1', item, 7.6. Further the periodic testing of.the area monitors and criticality detector as reworded. The current.
requirement to check the. channelis now a test.using a radiation source.At least monthly test are required, even if the reactor is not operating. The associated basis was reworded for clarityl .
: 70.      Annual calibration of portable instruments adde0das a surveillance and
       -explanation added to Bases.
: 71.      Ref. 2, Section 5, was reorganized to agree with the brganizatio'n&#xfd; arid section titling of ANS-15.1-2007. Specifically, paragraphs 5.1, 5.2, and 5.3. combined into a new. 5.1. Paragraph 5.4.1 and 5.5 cpmbined as new 5.2 with minor
* rewording in the first paragraph. Paragraphs 5.4.2, 5.4.3 and 5.4.4 w;ere combined in new 5.3. The first paragraph of 5.6 is renumbered as 5.4 and the remaining paragraphs were moved to the new Section 3.9 as discussed previously. The last sentence of the new 5.4 was slightly revised.
: 72. Ref. 2, Section 6 was rewritten in conjunction with a revised Safety Analysis Report, Chapter 12..
: 73. Ref. 2, paragraph 6.1.1 was revised to agree with the structure recommended by ANS-15.1-2007. Some wording changes were also made to show reporting relationships and the responsibilities were moved to 6.2.2. The paragraph describing the health physicist's relationship with the facility organization was
    ".reworded to include the Facility Director, moved to 6.2.2,1 an'd the position
      ,identified as Radiatioh Safety Officer.-
: 74. Responsibility of the Facilify Director andthe Operations Supervisor Were reworded.
: 75. Ref. 2, paragraph 6.1.3(a)3) and 4) combined.
: 76. Ref. 2, paragraph 6.1.3(c)1) revised to allow fuel and control rod.relocations in accordance with an approved experiment without direction from the Operations Supervisor.
: 77. Minimal staffing requirement added to Ref. 2, paragraph 6.3.1 for the condition of reactor shutdown, but not secured, as proposed by Ref. 1, 12.5. This inserted paragraph pushed the following paragraphs back one step.
: 78. Updated a reference in Ref. 2, paragraph 6.1.4 as proposed by Ref. 1, 12.7.
: 79. Ref. 2, paragraph 6.1.5 renumbered to section 6.2. This causes renumbering of subsequent sections.
: 80. Ref. 2, paragraphs 6.1.5.1 and 6.1.5.2 revised to refer to the NSRB Charter for its composition.
: 81. Ref. 2, paragraph 6.1.5.3 revised by adding item (d) for specific procedures requiring approval.
: 82. Added a new section 6.4, Radiation Safety.
: 83. Ref. 2, sections 6.2 through 6.6 renumbered as 6.4 through 6.8 respectively.
: 84. Ref. 2, section 6.3 1) completely rewritten as proposed by Ref. 1, 12.15.
: 85. Ref. 2, Section 6.4.1 was deleted since there is no safety limit. Section 6.4.2 was moved up to become section 6.4, then renumbered at 6.6.
: 86. Ref. 2, section 6.4.2(a) (now renumbered to 6.6 1) ) revised to require reactor shutdown as discussed in response to Ref. 1, 12.17.
: 87. References to 10CFR Part 50.59 in Ref. 2, section 6.5.1 revised as proposed in Ref. 1, 12.18.
: 88. Health Physicist changed to RSO in Ref. 2, 6.5.1 (a) 6) c) for consistency with other parts of the Technical Specifications.
: 89. Ref. 2; 6.5. 1 (d) revisedi to include preventative maintenance as proposed by Ref.
: 90. Ref. 2, 6:5.1(e) "revisedto clarify that license modificatio'ns' are required only of a more severe accidentis identifiedi
: 91. Ref 2, 6.15.1 (h)'rev'ised to'inchide visitors as pr'opose'dby Ref. 1;-12.18.
: 92. Reports in Ref. 2, 6.5.2 reworded as proposed by Ref. 1, 12.19 and a report added 'for changes to, evel 1 or Level 2 personnel. Refenre ce to telegraph reports was deleted since phone repoas to the NRC Operations Center can be made at any time.                    -              ..
: 93. Ref. 2, section 6.6 completely reyised to agree, with ANS-15.1-2007 in types of records to be retained and minimum retention time as proposed by Ref. 1, 12.20.
: 94. Ref 2, section 6.&. 1,revised to show.the entire name of the university.}}