ML19246A542

App a to Tech Specs,Reflecting Amend 14 Change Permitting Power Increase

ML19246A542
Person / Time
Site:	05000054
Issue date:	05/17/1979
From:	Office of Nuclear Reactor Regulation
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ML19246A533	List: ML19246A532 ML19246A540 ML19246A541 ML19246A542
References
NUDOCS 7907030271
Download: ML19246A542 (58)

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{{#Wiki_filter:. UN ION CARB1DE RESEARCH REACT 0R ~ TECHNICAL SPEC 1 FiCAT10NS APPENDIX A, License No. R-81 Docket No. 50-54 79070302m Amendment No. I4 }[J lj3 r Dated: May 17,1979

TABLE OF CONTENTS Page 1.0 DEFINIT 10NS.......................................................... I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS.................... 4 2.1 - Sa fe ty L i mi t s of Reac tor Ope ra t i on............................. 4

2. 2 - L i mi t i ng Sa f e ty Sys tem Se t t i n gs................................

7 3.0 L I M I T I N G CO N D I T I ON S FO R O P E RAT 10N.................................... 9 3 1 - Reactivity Limitations......................................... 9 3. 2 - C on t r o l a n d S a f e t y S y s t e ms..................................... 10 3.3 - Radiation Monitoring Systems................................... 13

3. 4 - Eng i n ee red Sa fe ty Fea tu re s.....................................

14 3.5 - Limitations on Experiments..................................... 15 3.6 - Fuel........................................................... 20 3.7 - Pool Water Quality............................................. 22 3.8 - Radioactive Releases (Airborne)................................ 22 3.9 - Rad iolog i ca l Envi ronmen ta l Mon i torin g.......................... 26 3.10 - La n d U s e C e n s u s................................................ 26 3.11 - Bases for Envi ronmen ta l Spec i f i ca t ions......................... 26 4.0 SURVEILLANCE REQUIREMENTS............................................ 31 4.1 - General........................................................ 31

4. 2 - S a f e ty C h a nn e l Ca l i b ra t i on.....................................

31 4.3 - Reactivity Surveillance........................................ 31

4. 4 - Con t rol and Sa fe ty Sys tem Su rve i l l ance.........................

32

4. 5 - Rad i a t i on Mon i to r ing Sys tem......................

32

4. 6 - En g i n ee red Sa fe ty Fea t u re s.....................................

32 4.7-ReactorFuel................................................... 33 4.3 - Sealed Sources................................................. 33 4.9 - Pool Water..................................... 33 4.10-CoreSpray..................................................... 34 4.11 - Flux Distribution... 14 5.0 DESIGN FEATURES.................... 34 5.1 - Reactor Fuel........... 34 5.2 - Con trol and Safety Sys tems.. 35 5.3 - Rod Control System.. 37 i llJ l36 C ~

TABLE OF CONTENTS (Cont'd.) Page 38 5.4 - Cooling System................................................. 5.5 - Containment System............................................. 39 41 5.6 - Fuel Storage................................................... 41 6.0 A D M I N I S T RAT I V E C O N T R 0 L S.............................................. 41 6.1 - Organization................................................... 6.2 - Procedures..................................................... 47 48

6. 3 - Expe r i me n t Re v i ew a nd Ap p rova l.................................

49 6.4 - Required Actions..... 6.5 - Reports........................................................ 50 6.6 - Records........................................................ 53 70 REFERENCES........................................................... 55 ii J / j'

1. l.C DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Condition of Operation (LCO) are as defined in 50.36 of 10 CFR Part 50. 1.1 Safety Channel - A Safety Channel is a measuring or protective channel in the reactor safety system. 1.2 Reactor Safety System - The Reactor Safety System is a combination of safety channels and associated circuitry which forms the auto-matic protective system for the reactor, or provides information which requires the initiation of manual protective action. 1.3 Operable - Operable means a component or system is capable of' performing its intended function in its required manner. 1.4 Ooerating - Operating means a component or system is performing its intended function in i ts normal manner. 1.5 Channel Check - A Channel Check is a qualitative verification of acceptable performance by observation of channei behavior. 1.6 Channel Test - A Channel Test is the introduction of a calibra-tion or test signal into the channel to verify that it responds in the specific manner. 1.7 Cha inel Calibration - A Channel Calibration is an adjustment of the channel components such that its output responds, within speci fied range and accuracy, to known values of the parameter which the channal measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip. l.8 Unscheduled Shutdown - An Unscheduled Shutdown is any unplanned shutdown of the reactor, af ter startup has been initiated. 1.0 Reactor Shutdown - The reactor is shut down when the negative reactivity of the cold, clean core including the reactivity worths of all erperiments is equal to or zreater than the shutdown margin. 1.10 Reactor Ocerating - The reactor is considered to be operating whenever it is not secured nor shutdown. hflJ l38 r

2. 1.11 Reactor Secured - The reactor is secured when: a. The core contains insufficient fuei to attain criticality under optimum conditions of moderation and reflection, or b. The moderator has been removed, or c. (1) Minimum number of control rods fully inserted as required by Technical Specifications, and (2) The console key switch is in the off position and the key is removed f rom the lock, and (3) No work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods, and (4) No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller. 1.12 True Value - The True Value of a parameter is its actual value at any instant. 1.13 Measured Value - The Measured Value of a parameter is as it appears on the output of a measuring channel. 1.14 Measuring Channel - A Measuring Channel is the combination of sensor, lines, ampli fiers, and output devices which are connected for the purpose of measuring the value of a parameter. 1.15 Recartable Occurrence - A Reportable Occurrence is any of those cord'tions described in Section 6.5.3 of this specification. 1.16 An Exceriment - An Experiment is an apparatus, device or material, placed in the reactor core, in an experiment facility, or in line with a beam of radiation emanating from the reactor, excluding devices designed to measure reactor characteristics such as detec-tors and foils. a. Secured Experiment - Any expe rimen t, experi ment faci li ty, or compcnent of an experiment is deemed to be secured, hfls 139 r

3 or in a secured position, if it is held in a stationary position relative to the reactor core. The restraining forces must be substantially greater tha7 those to which the experiment might be subjected by hydraulic, pneumatic, or other forces which are normal to the operating environ-ment of the expe rimen t (or by forces which can arise as a result of credible malfunctions). Movable Experiment - A movable experiment is one which may be inserted, removed, or manipulated while the re-actor is critical. c. Untried Exoeriment - is a single experiment or class of experiments that has not been previously evaluatec and approved by the Nuclear Safeguards Committee. 1.17 Experiment Facilities - An Experiment Facility is any structure, device or pipe system which is intended to guide, orient, posi-tion, manipulate, control the envi ronment or otherwise facili tate a multiplicity of experiments of similar character. 1.18 Control Rod - A control rod is a rod fabricated f rom neutron absorb- 'ng material which is used to compensate for fuel burnup, tempera-ture, and poison effects. A control rod is magnetically coupled to i t-drive unit allowing it to perform the safety function when the magnet is de energized. 1.19 Readily Available on Call - Readily available on call means an individual who, (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty. 'ormed of where he may be rapidly contacted (e.g. by phone, etc.) (3) is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g., I hr. or within a 30 mile radius). 1.20 Scran Time - is the elapsed time between the instant a limiting safety system set point is reached and tne instant that the s lowes t control rod is fully inserted. fo /40 E

4. 5 fety Limits - are limits on important process variables which are 1.21 found to be necessary to reasonably protect the integrity of certcin physical barriers which guard against release of radioactivity. The principal physical barrier is the fuel cladding. 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits of Reactor Operation 2.1.1 Limits in Forced Cooling Mode a. Aeolicability - This specification applies to the variables that affect thermal and hydraulic performance of the core during forced cooling. They are: (1) Power in MW. (2) Flow in GPM. (3) Height of water above the core, b. Objective - To assure fuel cladding integrity, c. Soeci fications - (1) The maximum steady power level under various flow conditions shall be as shown in Figure 1. (2) The pool water level shall not be less than 20 feet above the core. d. Bases - The analysis given in Ref.1, Sec. A1, forms the basis for this specification. The superposition method of Gambill is used to derive the burnout heat flux as a function of primary flow rate. A safety factor of 1.25 is applied to allow for uncertainties in the corre-lation. Pool temperature (or core inlet temperature) is not included in the specification as this variable changes very slowly and has only a minor effect, e.g., a 10*F change results in only a 5% variation in burnout flux. The lat ter, howeve r, is evaluated conservatively near the high end of the pool temperature range that is ex-pected in practice. A de-rating factor can be applied for pool temperatures in excess of 120*F. The relation-ship between total power and peak heat flux is derived 5 14l

5. for the core situation with the greatest peaking factors, viz. a new fuel element adjacent to a central in-core flux trap. Reactor power, primary flow rate, and water level will be maintained well within safety limit speci-fications through limiting safety system scram settings. (see 2.2.1). 275 pp

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7 2.1.2 Limits in Free Convection Mode a. Aoplicability - This specification applies to the thermal and hydraulic variables af fecting the core during natural con-vection cooling. They are: (1) Power in MW. (2) Height of water above the core. b. Objective - To assure fuer cladding integrity. c. Specifications - (1) The maximum reactor power level shall be 6.7 MW. (2) The pool water level shall not be less than 20 ft. above the top of the core. d. Bases - The analysis given in Ref.1, Sect. A2, forms the basis of this specification. The homogeneous method of Gambill and Bundy, used in this analysis, has been employed successfully to predict natural convection burn-out in ORR and HFIR fuel. The fe rmer fuel i s close in design to UCRR fuel. A safety factor gf.l.24 is applied to account for random variations and uncertainties. A pool temperature near the high end of the operating range (120*F) is assumed. The safety system settings on power and pool level (2.2.2) assure adherence to these speci-fications. 2.2 Limiting Safety System Settings 2.2.1 Safety Channel Set-Points in Forced Cooling Mode a. Acolicability - This speci fication applies to the set-points of the safety channels. b. Objective - To insure that automatic action is initiated that will prevent a safety limi t from being exceeded. c. Soeci fication - For cperation in the forced cooling mode the limi ting safety system settings are: (1) Powe r level at any flow rate shall not exceed 7.5 MW. 275 144

8. (2) Power level settings for vari,ous conditions of flow and of pool temperature shall be in accord-ance with Fig. 1. (3) Coolant flow shall not be less than 1800 gpm for powers above 250 KW, (4) Pool level shall not be less than 20 ft. above the top of the core. d. Bases - Safety limits have been shown previously (Sec. 2.1 and Ref.1) to lie at a low flow-to power ratio. To pro-vide adequate assurance that these limits are not approached too closely, the LSSS are chosen conservatively so as to minimize the chance of boiling in the core. This results in a much larger flow / power ratio. In Ref. 1, Sect. A3, power levels derived using conservative correlations for incipient boiling are tabulated for various values of pool temperatures and flow rates to illustrate the result-Ing temperature margins. Through a comparison wi th experi-ments at ORNL (ORR) this method is shown to be conserva-tive. To preserve the desired temperature margins for all combinations of variables, the LSS settings are a com-bination of two fixed set-points, viz. scrams at 7.5 MW and 1800 gpm, plus an adjustable one that provides auto-matic power reduction at a setting that depends on the pool temperature and flow rate. Rates of change of pool temperature are very s low - a few 'F/hr. at most - and thus allow adequate lead time for adjustment. For a reactivity transient the case considered is the step insertion of 0.25 % a K positive reactivits with the reactor operating at a steady power of 7.5 MW. The analysis given in Ref. 1, Sect. B3, shows that the pcwer at the end of.75 sec. (the scram time, Sect. 3.2.1 below) will be no more than 11 MW. This is well below the safety limit for this mode of operation. fs E /ig L

9 No autyatic scram is associated wi th pool temperature as t'- , parameter varies very slowly allowing ample time f appropriate operator action. 2.2.2 Safety Channel Set-Points in Natural Convection Mode a. Apolicability - This specification applies to the set-points of the safety channels. b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded. c. Scecification - For operation in the natural convection mode, the limiting safety system settings are: (1) Power Level I 250 KW. (2) Pool Level 3 20 ft. above the core. d. Bases - The set points are chosen to avoid boiling in the core during routine operation with natural convec-tion cooling. The analysis given in Re f. 1, Sect. A4, shows that a power of 0.35 MW is needed for incipient boiling to occur. To allow for uncertainties a safety factor of 1.3 is applied to this, resulting in a safety system set point of 0.25 MW. The latter is well below the safety limit of 6.7 MW given above (Sect. 2.1.2). In the case of reactivity transient, a step insertion of 0.25% a K positive reactivity at an initial power level of 0.25 MW will, following the analysis of Ref.1 (Sect. B3), resul t in a transient power of 0.38 MW af ter 1 second. The latter is well below the safety limit of 6 7 MW for the natural convection mode (2.1.0). 3.0 LIMITING CCNDITICNS FOR OPERATION 3.1 Reactivity Limitations 3.1.1 Shutdown Margin The minimum shutdown eargin provided by control rods in the cold, xenon-f ree ccndi t i on wi th the highes t worth rod fully withdrawn 27oc /

10. i and with the highest-worth non-secured experiment in its nost posi-tive reactive state shall not be less than 0.5% a K. This specification ensures that the reactor can be shut down from any operating condition and remain shut down after codl-down and xenon decay even if the highest-worth control rod is stuck in its fully withdrawn condition. 3.1.2 Excess Reactivity The core shall not be loaded with an excess reactivity of greatet than 10.2% a K when located in the stall position and 8.2% a K when the core is located in the open pool position. 3.1.3 Expe ri men ts Reactivity limits on experiments are specified in 3 5 below. 3 1.4 Regulating Rod The integral worth of the regulating rod shall not exceed 0.6% a K. This ensures that a malfunction of the control system cannot make the reactor prompt cri tical. 32 Control and Safety Systems 3.2.1 ,5 cram Tim.e The scram tire shall not exceed.75 second and the control-rod magnet release time shall not exceed.05 second. In the transient analysis (Ref. 1, Sect. B3), these values were assumed. 3.2.2 Measuring Channels The minimum number and type of measuring channels operable and providing information to the control room operator required for reactor operaticn are given as follows: Channel No. Ope rab l e_ Operating Mode in Which Recuired Power Level (no rma l ) 2 All Power Level (intermediate) 1 Aj1 Period Channel 1 All fa 147 E

11. Channel No. Operable Operating Mode in which Recui red Count Rate 1* Startup Coolant Flow 1 Forced Cooling Core a T 1 Forced Cooling Rod Position 1/ rod All Pool Temperature 1 All Pool Level 1 All Note: a. Operable below 50 W. Bases - The normal power level instruments (" Level Safe-ties") provide redundant information on reactor power in the range 25%-150% of the normal operating power level of 5 MW. The intermediate power level instrument (" Log N") provides usable reactor power Information in the logarithmic range 4 10 %-300% of the normal power of 5 MW. The count rate channel covers the neutron flux range from the source level ($ 1 cps) to 10 cps on a logari thmic scale. It enables the operator to start the reactor safely from a shutdown condi tion, and to bring the power to a level that can be measured by the Log N instrument. Ccolant flow rate and a T instruments allow the operator to calculate reactor power and calibrate the neutron flux channels in te rms of power. Rod position indicators show the ope rator the relative positions of control rods, and enable rod reactivi ty calibrations to be made. Pool temperature information allows the operator to adjust the cooling system to keep nool terperature within a pre-ferred range, and to adjust the ove rpowe r reverse set-point (see 3.2.3). 27'c 148

12. 3 2.3 Safety Channels The minimum number and type of channels providing automatic action that are requi red for reactor operation are as follows: Channel No. Operable Function Operating Mode Power Level (normal) 2d Scram ? > 7.5 MW All Power Level (Intermediate) 1 Scram 3 < 3 sec. period All Reverse @ < 10 sec. period All inhibit 9 < 30 sec. period All c All Reverse Count Rate 1 inhibi t @ < 2 cps Startup 8 inhibit ? < 30 sec. period Startup Pool Water Level 1 Scram @ < 22 ft. All Pool Temperature 1 Alarm @ > 120*F All Coolant Flow I Scram @ < 1800 gpm Forced Cire. Manual Button i Scram All Bridge Lock 1 Scram All Guide Tube Lift 1/ rod Scram All Flapper Valve 1 Scram (above 250 kW w/ valve open) All Keyswitch i Scram All Notes: a. Operable below 50 W. b. Operable above 250 KW. Overpower reverse set points shall 8,e set so that the c. relationship of pool temperature, flow and power levels shown in Figure I are never exceeded, d. Two power level (normal) channels shall be required to be operable for reactor operation. A third uncomp. nsated ion chamber is considered a backup shall be operaole prior to reactor startuo. Bases - The power level scram provides redundant auto-ratic protective action to preve.it exceeding the safety l i r.i t on reactor powe r. The period scram, assisted by the i n te rmed i a te level period reverse and rod inhibit, limits the rate of increase in reactor power to values that are controll-able without reaching excessive power levels or tempera-ture. These functions are not limi ting safety system settings. C 149 a

13 The two inhibits on the count rate channel prevent inad-vertent criticality during cold startup that could arise f rom lack of neutron information or f rom too rapid re-activity insertion by control rods. lhe scram ca pool level provides an adequate head of w6ter above the core and guards against loss of coolant and loss of building containment. The overpower reverse nn the intermediate power channel provides automatic action to reduce power and minimize the chance of incipient boiling in the core. The coolant flow and flapper valve scrams ensure adequate coolant flow to prevent boiling in the core. The scrams on bridge lock and guide tubes prevent unplanned reactivity changes that could occur through core and con-trol element movements respectively. The keyswitch scram prevents unauthorized operation of the reactor. Bypass is permi tted on those parameters that can be moni-tored by alternate means if the initiating circuit mal-functions. 33 Radiation Monitoring Systems The mininum acceptable nonitoring instrumentation required for reactor operation is as follows: h Alarn Funct. ion Type No. Operable Setuoint Excursion Monitor 1 5R/hr Detect high radiation: Alarm and isolate at > 5R/hr. Exhaust Duct Monitor 1 Detect particulate, gas and (" Stack Monitor") iodine activities; alarm in Control room. Building CAM 1 Detect particulate activity in reactor building; alarm. The alarm set point for the stack gas monitor shall not be set above a value that would result in an ex;osure greater than 2 mrem / hour assuming a dilution factor of 2000 and the isotope mixture determined by the most recent analysis. The alarm set point for the stack I-131 and stack particulate monitor shall not be set above a value corresponding to that listed in Aopendix B, Table II, Column I of 10 CFR Part 2C assuming a dilution factor of 2000 and averaging over one week.

• • 25; of the maximum permissible concentration at restrictedhaf f to Appendix B of 10 CFR 20.

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I in. I Max. Alarm i Type No. 00 arable Setcoint Function Fixed Area Monitron 3 50mr/hr Detect radiation (y) in key locations; alarm in Control room. Evacuation Swi tch 1 Alarm and initiate evacuation sequence. (manual). Note: For maintenance or repair, required radiation monitors (except for excursion moni tor) .?y be replaced bv portable or subscitute instruments for periodc up to 24 hours pro-vided the fu.ct.on will still be accomplished. Interruption for brief periods to permit checking or calibration is per-missible. 3.4 Engineered Safety Features These specifications apply to required equipment for the confine-ment of activity through controlled release of reactor building air to the atmosphere. 3.4.1 Excursion Moni tor a. Soecification: see 3.3 b. Basis - This monitor senses excessive radiation at the reactor bridge and autcmatically initiates the "evacua-tion sequence", whicn consists of a distinctive alarm, closure of damper valves in the building ventilation system and hold-up tank vent, and starting of the emer-gency exhaust fan (see 5. 5.2). 3.4.2 Emercency Electric Generator S aec i fi ca r i_gn a. Eaui;, _gt No. Ooerable Function Electric Generator 1 Upon loss of utili ty power, start aute atically and supply eme. gency cower to the exhaust fan and ventilation system controls. A six day supply of fuel giall be main-tained. ()r /

'5 b. Basis - Upon loss cf utility power the reactor scrams autoratically. Controlled release confinement reauires the ability to run the emergency exhaust fan ani to close build:ng damper valves. The latter are pneumatically-operated but are electrically-controlled. 3.4.3 Con ta i nme n t a. Soeci fi ca t i on (1) The emergency exhaust fan shall be capable of sustaining a negative pressure within the re-actor building of at least.01-in w.g. at an exhaust flow rate of not greater than 200 cfm. (2) Filters in the emergency exhoust shall be HEPA and charcoal with tested efficiencies of 99.5% for particle removal and 95% for iodine removal respectively. (3) Depth of water in the canal shall be at least 10 ft. This is equivalent to a water height above the core of 22 ft. (4) At least one door of the double airlock doors and the truck doors shall be closed while the reactor is operating. b. Bases - To effect controlled release under accident con-ditions of gaseous activity present in the building at-mosphere, a negative pressure is required so that any building leakage will be inward. Reference 1 (Sec. C, 2) contains an analysis of a hypothetical accident resulting in release of ai rborne activi ty to unrestricted areas. The assumed exhaust rate is 200 cfm and the filter effi-ciency for elemental iodi,e is 95%. In the design of the containment building (5.5) tne water seal in the canal is effected when the water depth is > 10 ft. 3.5 Limi tations on Expe r i men t s [I Jq2

'6. 3.5.1 Experiments a. Apolicability - This specification applies to those experi-ments installed in the reactor and its experiment facilities. b. Objective - The objective is to prevent damage to the reactor or excessive release of raaloactive material in the event of an experiment failure and also to prevent the safety limits from being exceeded. c. Specification - Experiments installed in the reactor shall mee t the following conditions: (1) The combined worth of all experiments which can add positive reactivity to the core due to a com?on-mode failure shall not exceed 2P.57. (2) The combined worth of all non-secured experiments which can add positive reactivity t6 the core due to a common-mode failure shall not exceed 1.7%aK. (3) The reactivity of any single experiment shall not exceed 0.5% a K. (4) An experiment worth less than 0.25% a K may moved when the reactor is critical. (5) An experiment worth more than 025% a K but less than 0.5% a K may be moved with the reactor sub-critical by at least 0.75% a K. (6) All material to be irradiated in the reactor shall either be corrosion resistant or encap-sulated within corrosion-resistant containers. (7) Where failure of the pressure-containing walls of an experiment container can cause a hazard to personnel or to the reactor, the container shall be designed and testad in accordance with t.ie applicable pressure vessel codes. (3) in-core experiments exposed to reactor water shall be designed to prevent surface boiling. 275 753

17 (9) Experimental apparatus, material, or equipment to be inserted in the reactor shall not inter-fere with the safe operation of the reactor. (10) The total primary coolant flow utilized by all in-core experiments shall be limited to the same as that in six standard fuel elements. (11) Experiments on the grid plate extension are limited to a total reactivity of 0.2% 6 K and a total load of 100 lbs. (12) Each class of experiment irradiated in the reactor must have been previously reviewed and approved by the Nuclecr Safeguards Committee (6. 8). d. Bases (1) See Ref. 2, Sect. G6. (2) It is shown in Ref. 3 that the reactor can safely self-limit a step reactivity insertion of $2.14. This corresponds to an insertion of 2.14 x.81 = 1.73% a K. (3) The method of Ref.1, Sect. 83, shows that a step insertion of 0.5% a K with the reactor critical at 5 MW (or 0.25 MW, in natural con-vection mode) will result at the end of.75 sec. In a power of not more than 14 MW and. 4 ltW, for natural convection. Each of these power levels does not exceed the corresponding safety limit. (4) Similarly it is shown that a step increase of 0.25% a K will produce a power level at the end of the scram time that is much less than the safety limit in ei ther mode of operation. In addition.25% a K is well within the auto-matic control cacabili ty of the reactor cont-rol systen. 27c 1:74 e g

18. (5) This specification ensures that, even with a 45% error in estimation of the reactivity of an experiement, the reactor will not be made critical. Even if the reactor were cri tical, the resulting period (% 3 secs.) will automatically initiate corrective control action. (6) This requirement guards against release of acti-vation products in the primary coolant or chemi-cal interaction with core components. (7) This specification ensures that there will be no mechanical damage to the reactor core nor hazards to personnel due to failure of experiment con-tainers where pressure exists or builds up dur-Ing irradiation. In the case of fueled experi-ments, it further ensures against hazardous and uncontroll.d release of fission products into the reactor building or the envi ronment from the same cause. (8-9) Ensures that no physical or nuclear interference with the safe operation of the reactor will occur. (10) This ccndition is assumed in the analysis given in Ref. 1, Sect. A. (11) These limits ensure that movement of these ex-periments will not result in reactivi ty changes in excess of tiiat in Sect. c(4) above. (12) Ensures that all experiments are evaluated by an independent group knowledgeable in the appropri-ate fields. 3.5.2 Fueled Exceriments a. Aoplicabili ty - These specifications apply to experiments containing nuclear fuel that are installed in the reactor or its experiment facilities. f'C 15o n

19 b. Objectives - The objective is to prevent damage to the reactor, prevent excessive release of fissien products in the event of an experiment failure, and also to en-sure that safety limits are not exceeded. c. Speci fications - Fuel-bea ring experiments in the reactor shall meet the following conditions: (1) All fueled experiments are to conform to the specifications listed above in Sect. 3.5.1. (2) The inventory of solid fuel bearing material being irradiated in the reactor core at any time shall be limited to 200 g of source and/or 750 g of special nuclear material. (3) The inventory of solid fuel-bearing materials in a single irradiation capsule shall be limited to 200 g of source and/or 50 g of special nuclear material. (4) The fission power of an irradiation capsule containing special nuclear material shall be limited to 13 KW. (5) The iodine inventory of a single capsule shall 131 be limited to 500 curies I dose equivalent for a doubly-encapsulated capsule and 70 curies 131 I dose equivalent for a singly encapsulated capsule. d. Bases - These specifications place limits on the fission product inventory in a fueled capsule such that capsule failure and the hypothetical release of all contained fission products to the reactor coolant will not result in excessive exposure to personnel on and of f si te. The detailed analyses that form the bases of this speci-fication are given in Ref. 1, Sec. C 3 The total amount of special nuclear material permi tted in the core at any time has been increased to 750 g. This increase does not 27'c !!i6

20. affect the consequences of the release from a single capsule as analyzed in Ref. I for it has been estab-lished (see Li cense Amendment No. 10) that failure of a single capsule will not initiate failure in other neighboring capsules. The core limit of 750 g is based on approximately 15 irradiation positions, each holding 50 g of SNM. The limit of 15 positions is dictated by availability of primary cooling capacity. 3.6 Fuel a. Applicability - These specifications apply to the number and condition of the fuel elements present in the core. b. Objective - To ensure that power is distributed in the core among a sufficient number of fuel elements to avoid excessive peak / average ratio, and to avoid ex-cessive release of fission products. c. Specifications (1) The minimum number of fuel elenents in the core shall be 30 Each control element shall count as 1/2 fuel element for this purpose. (2) Control rods shall be kept within I 10% of their mean position whenever the reactor power exceeds 500 KV. (3) Fuel elements exhibi ting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. An increase in the normal gaseous fis-sion product release (due to system contami-nation) by a factor of 100 shall consti tute initial evidence of cladding rupture and require identification of the cause. (4) Fuel element loading ari Jistribution in the core shall be such that peak / average thermal flux will not exceed 3.3 E 157

21. (5) The fuel plates are composed of enriched uranium-aluminum sandwiched between high pur-ity aluninum clad. Fuel plates may be fab-ricated by alloying the uraniem-aluminum or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in the fine powder form. Burnup of the fuel assemblies shall be limited to.c4 x 10 ' fission /cm '. Fuel plates may also be fabri-cated from uranium oxide-aluminum (U 0 -A1) 33 using the powder metallurgy process and the 2I burnup shall be limi ted to 1.5 x 10 fission / 3 cm. d. Bases (1) A minimum of 30 elements is assumed in the analysis given in Ref. 1, Sects. A 1, A 2. (2) This specification minimizes flux tilts that could cause concentrations or shif ts in power distribution across the core. Such shifts are only signi ficant in power operation, and thus this limitation is restricted to power levels above 10% of the normal 5 MW. (3) Release of fission products from the compact fuel plates used in this reactor (Sect. 5.1) due to a localized cladding defect is con-fined to the defect locality. A relatively small de fect thus cannot release large quan-tities of fission products. There is a normal small and variable background of fission product release due to uranium contamination in the coolant and on fuel plates. It i s thus safe to spe'.ify a recognizable and substantial increase in this background as a possible indication of cladding rupture. If the n I 76/ J

22. runture were extensive, there would be no doubt at all of this condition. (4) This peak / average value is used in the Ref. I analysis. (5) Amendment No. 12 and Ref. 6. 3.7 Pool Water Quality a. Apolicability - This specification applies to primary cooling system water in contact with fuel elements, b. Objective - To minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials. c. Soec i fi ca t i ons (1) Pool water temperature will not exceed 130*F. (2) Pool water specific resistance is to be not less than 200,000 ohm-cm, except that for periods not greater than 14 days it may be > 70,000 ohm-cm. (3) The pH of the pool water shall normally be maintained between 5.0 and 7.5 3.8 Radioactive Releases 3.8.1 Ai rborne Stack Release Limi t Maximum yearly release rates for noble gases, radioiodines and particulates of half-life greater than eight days shall be limited by the following expression: ? Q. (x/Q)/MPC. < 1/6 i e i where: Q; = The average release rate for any 12-consecutive months of radionuclide, i, in gaseous effluent from the stack in Ci/sec. MPC; = Activi ty concentration of radienuclide, i, as given in Table !!, Column 1 of Appendix B to 10 CFR 20, in a Ci/cc. 27'c 1:79

23 = Shall be calculated monthly from measured values X/Q of iodine concentration sampled at or above the tree line 380 meters NE of the exhaust stack. ACTION: Should the limit of this Section be exceeded.the licensee shall notify the Commission within 24 hours, and take action to reduce the release to within the limi ts immediately. 3.8.2 Jose in Unrest'ricted Areas a. Total body dose due to noble gases releases and dose from radio-iodines in gaseous effluents for the critical individua s in unrestricted areas should be calculated at least once per calendar quarter and reported in the annual report (TS 6.5.2g) b. The total body dose to any individual in unrestricted areas due to noble gases released in gaseous effluents from the site shall be limited to the following expres-sions: i) During any calendar quarter: -8 3 17 x 10 M; (X/Q) Q; 5 2.5 mrem 'I) During any calendar year: -8 3 17 X 10 { M; (X/Q) Q; - 5 mrem where: Q; = The release of noble gas radionuclide, i, (measured concentration x flow rate) in.aC1. Releases shall be cumlative over tne calendar quarter or year as appropriata. M = The total body dose factors due to gamma emissions for each identified noble gas radionuclide, mrad / year per uCi/m from Table B-1 of Rev. 1, Reg. f- [6g Guide 1.109

24. X/Q = Shall be calculated from measured values of iodine concentration samoled at the environ-mental monitoring station in Laurel Ridge. This measured value shall be increased by a factor of 2 when calculating the body dose limits. ACTION: With the calculated air dose from radioactive noble gases in gaseous effluent exceeding any of the above limits, prepare and submit to the Commission and New York State Departrent of Environmental Conserva-tion, within 30 days, a special report which identi-fles the cause(s) for exceeding the limit (s) and defines corrective actions to be taken to reduce the releases. c. The dose to an individual from radioiodines in gaseous ef fluents released to unrestricted areas shall be limited to the following expression: i) During any calendar quarter: 3.17 x 10~ I (R; W Q ) 5 7.5 mrem, and i li) During any calendar year: -8 3.17 x 10 I (R. W Q ) - 15 mrem e i where: Q. = The release of radiciodines in gaseous ef fluents, I i, in uCi. Release shall be cumulative over the calendar quarter or y ear, as appropria te. 27c 16/

25. R; = The dose factor for each radioiodine, i, in mrem /yr per uCi/m3 (from Reg. Guide 1.109) except for I-125 which is detemined tn he as follows: Adult Thyroid (inhalation) - 1.1 x 10-3 mrem /pCi Infant Thyroid (inhalation) - 6.8 x 10-3 mrem /pCi (ingestion) - 8.9 x in.3 mrem /pCi W = The average dispersion parameter for estimat-ing the dose to an individual in the controlling location from radiciodines in gaseous effluents re-leased to unrestricted areas. W = (x/Q) for the inha:ation pathway, in sec/m (as determined in 3.8.2.a) ~ W = (D/Q) for the food pathways, in meters ACTION: With the calculated dose from the release of radio-iodines exceeding any of the above limi ts, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, c Special Report which identifies the cause(s) fc. exceeding the limit and defines the corrective actions to be taken to reduce the releases. NOTE: The present controlling dose pathway is via infant inhalation at the Laurel Ridge Residential site. If the Land Use Census (Section 3.10 of this speci-fication) identifies a location or pathway which yields a calculated dose or dose corvnitment greater than via the presently calculated dose pathway, the dispersion parameter (x/Q or D/Q) and dose factor (R;) for this more restrictive pathway shall be used in this specification. 27'c 162

26. 3.8.3 _ Liquid Effluent Releases a. Liquid waste from all radioactive operations shall be collected in hold tanks, b. Before release from the hcid tanks, the liquid waste shall be sanpled and the activity level measured. c. Liquid waste shall not be released from the site unless its activity concentration, including dilution with non-radioactive waste water, is below that specified in 10 CFR, Part 20, Appendix B Table II, Column 2. This activity concentration shall be determined at least once per month by an analysis of a composite samole of all tanks released duri.g that period. d. Total radioactivity released in liquid effluents shall not exceed 0.01 Ci (5r-90 equivalent) in any 12-consecutive month period. If the above limit is exceeded, make a special re-port to the NRC within 30 days explaining the cause of exceed-ing the limit and the corrective action to reduce the release to within the limit.

e. Records of and reports on liquid radioactive effluent releases shall be as specified in Section 6 of these Technical Specifications.

3.9 Radiolocical Environmental Monitorina The radiological environmental monitoring program shall be con-ducted as specified in Table 3.9.1. The results of analyses performed on the radiological environmental monitoring samples shall be summarized in an Annual Radiological Environmental Report. 3.10 Land Use Census A land use census shall be conducted at least once per 12 months between June 1st and Oct. 1st, and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh, leafy vegetables in each of the 16 meterological sectors within a dis-tance of five miles. Y fh l o j,

27 Ac t.l on With a land use census identifying a location (s) which yields a calculated dose or dose commi tment greater than at a location for which dose is currently being calculated in Specification 3.8.2.b and from which samples are currently,being obtained in accordance,with Specification 3.9. prepare and submit to the Commission and the' New York State Department of Envi ronmental Conservation, wi thin 30 days, a Special Report which identifies the new location. The new location shall be added to the radio-logical environmental nonitoring program within 30 days. The sampling locaticn having the lowest calculated dose or dose com-mitment (via the same exposure pathway) may be deleted from this monitoring program af ter (October 31) of the year in which this land use census was conducted. 3 11 Bases for Envi ronmental Speci fications Specification 3.8.1 is provided to ensure that the dose a. at the exclusion area boundary from gaseous effluents from tne site will be wi thin the annual dose limits of 10 CFR Part 20 for unrestricted areas. The speci fied release rate limits restrict, at all times, the cor-responding gamma and beta dose rates above background to an individual at or beyond the restricted area boundary to 1 (500) arem/ year to the total body. These release rate limi ts also restrict, at all times, the correspond-ing thyroid dose rate above background via the inhalation pathway to 1 1500 mrem / year. Specification 3.8.2 is provided to demonstrate compliance b. with 10 CFR 20.l(c) which requires releases of radioactive materials released to unrestricted areas to as low as rea-sonably achievable. The acticn statements provide the operating flexibili ty and at the same time implement the design objective of minimizing the release to unrestricted 27J !64 c

28. areas to as 1 74 5 reasenably achievable. The speci ri-cations for noble gas releases are based on limiting the total body dose at the limiting populated area to less than 5 millirem /yr. The specification for radiciodine is based on the assumption that the limiting dose path-way for these radioisotopes is via infant inhalation at the Laurel Ridge Residential Site, and limits the infant thyroid dose to less than 15 millirem /yr. c. The radiological monitoring program required by speci-fication 3.9 provides measurements of rc !!ation and radioactive materials in those exposure pathways and for those radienuclides, which lead to the nighest potential radiation exposures of individuals rer ting f rom the s ta t i on ope ra t i on. This inoni toring pc n thereby supplements the radiological ef fluent moni tor-ing program by verifying that the measurable concentra-tions of radioactive materials and levels of radiation are not higher than expected or the basis of the effluent measurements and modeling of the environmental exposure pathways. This monitoring program may change based on operational experience and results of the land use

census, d.

Specification 3.10 is provided to ensure that changes in the use of enrestricted areas are identified and that modi fications to the technical specification limit of dose via the most restrictive dcse pathway and the ntnitoring program can be made if required by the results of this census. 9 I c7J l63 e

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cza wr s g1 omps ey p p e l r oy7 yoooy r uool a a a i il tl il fl n>ti t ppd me me y T d ar radl r ae nd omm mc mc ane aneoenMsoes oa2 an an rap PA rfP avicmics9 G o G o M AR G O edt. R y eas rI l P c npr y 1 2 es n omit a 3 9 t o G e i auud ed l d u t s qb r r d ) f i nq a e 7 e et f R ae rh r g p pueod O r et nr or e T gF pi si e e e-aet I n owad p c cd pc n a n nasye N i o s rnoe o oeetl O l H pi ueol c Ri e 1 mt ol n t t( cys i 9 L ac upt t o s s nb nmcs a a e s A S e i aeut e. e .ud r 3 T l l t sl d s N l s l s qee E E o n l a y yent L 1M C of oye t a t arie B C ocbl Ad Adf mr A 0 T R. V i t E L A l C e I s s r e G e n 0 u v o 8 a o O l L p i 3 L b O m t f a I a a mo m. D S c o oa 1 A o rE re R f dL f l f r l on A s ae es. e a r lf_ l rk l e e pec pg e b m mt a md m m a aet ai a S ms SR S u S_ l i i t lr s 0 e e I y n t T a = i a A v! d l p o u I D t m l c A aa E o i R PS l i t i R dd r T e O ana C r r B R aP E uo R R s/ I I od A a D p n xa E I 2 ) p N(; g - sN - 7 3 - C C

TABLE 3.9.1 (Continued) RADIOLOGICAL ENVIR0tJMEliTAL M0tJITORiflG PROGRArt flumber of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locationn, Collection Frequency of Analysis 3. I t4 GEST l otl Food Products location to be At time of harvest.

• I-131 analysis.

determined from One sample of broad Land Use Census, leaf vegetation. Water Indian Kill inlet Monthly Gross Beta - Monthly Indian Kill outlet Monthly Gross Beta - 11onthly Warwick Brook Monthly Gross Beta Monthly Sterling Lake outlet Monthly Gross Beta - Monthly Ramapo River Monthly Gross Beta - Monthly -2 N

• The maximum values for the lower limit for I-131 are 7 x 10 Cl/m3 airborne N

concentration and 60 pCi/Kg, wet weight leafy vegetables. L7 w N

31. 4.0 SURVEILLANCI REQUIREMENTS 4.1 General The requirements listed below generally prescribe tests or in-spections to veri fy periodically that the performance of required systems is in accordance with specifications given above in sections 2 and 3 In all instr ices where the specified f requency is anr.ual, the interval between tests is not to exceed 14 months; when semiannua. e interval should not exceed 7 months; when monthly the interval shall not exceed 6 weeks; when weekly the interval shall not exceed 10 days; and when daily the interval shall not exceed 3 days. 4.2 Safety channel Calibrat tjn, A channel calibration of each safety channel shall be performed annually (see Se t. 3.7.3). 4.3 Reactivity Surveillan.:t (I) The reactivity. orth of each control od (including the regulating od) and the shut-down margin shall be determined w1enever operation requires a reevalua-tion of core p:ysics parameters, or annually, which-ever occurs fi rst. The rod worth will be determined using the reactivity-period or rod-drop methods. (2) The reactivity worth of an experiment shall be esti-mated, or measured at low power, before conducting the experimen t. (3) Boron /Carb!de rods shall be gauged quarterly and any dimensional changes reported promptly to the Com-mission. S i l ve r/ i nd i um/ Cadmium control rods shall be gauged annually, or, in the case of newly installed rods, at the end of the first six months, if any Ag/ In/Cd rod should be found not to meet the acceptance cri teria i t shall be removed f rom service. n addi-tion all other rods manufactured of the same batch shall be ir.spected. 275 168

32. 4.4 Control and Safety System Surveillance (1) The scram time shall be measured annually. If a control rod is removed f rom the core temporarily, or if a new rod is installed, its scram time shall be measured before reactor operation, if the bridge is moved, the scram time will be measured before subsequent reactor operation. (2) A channei test of each measuring channel in the reactor safety system shall be performed monthly or prior to each reactor operating period whichever occurs first unless the preceding shut-down period is 8 hours or less. A channel test before startup is, however, required on any channel receiving maintenance during the shut-down period. (3) A channel check of each measuring channel (except for the pool level) in the reactor safety system shall be per-formed daily when the reactor is in operation. 4.5 Radiation Monitoring System (1) The excursion, stack, and area monitors shall be cali-brated annually. (2) The excursion, stack, and area moni tors shall receive a channel test monthly. (3) The excursbn, stack, and area moni tors shall receive a channel check and a setpoint verification daily durina reactor operating periods. 4.6 Engineered Safety Features 4.6.1 Excursien Monitor: see above 4.5 4.6.2 Emergency Generator (1) The ability of t- , r: ene.ator to start, to run normally, and ancrote 420 VAC shall be checked weekly. 5 l69

33 (2) The generator shall be tested for its ability to accept, via the automatic transfer switch, the reactor electrical load once every six months. A commercial power outage and subsequent pickup of load by the emergency generator will count as a successful load test. 4.6.3 Containment (1) The efficiency of the charcoal filters and of the abso-lute filters in the emergency exhaust system shall be measured annually and the flow rate verified. (2) The operability of the evacuation alarm and containment isolation system shall be tested, ar.d negative pressure verified, semiannually. A utili ty power outage may be used to initiate such tests. 4.7 Reactor Fuel (1) Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifi-cations. (2) Each new fuel element will be inspected for damage and flow obstructions prior to insertion into the core. 4.8 Sealed Sources The antimony-beryllium sealed source shall be leak tested in accordance with the procedures described in the application for license amendment dated March 21, 1963, except that the frequency of leak testing will be in accordance wi th 10 CFP. Part 34.25(b). The strontium-90 sealed source shall be tested for leakage and/or contamination semiannually. 4.9 Pool Water (1) The pH and specific resistance of the pool water shall be determined each week. 275 170

34. (2) An analysis of the pool water for radioactive material shall be done at monthly intervals. This analysis is I to include Sb as an indicator of Sb-Be neutron source integrity. (3) Activity of the pool water will be measured weekly. 4.10 Core Spray The core spray in the reactor operating position shall be tested for operabili ty semiannually. 4.11 Flux Distribution in order to verify that power gradients amongst fuel elements do not cause peaking factors to exceed those used in the bases of Sect. 2.1, the radial neutron flux distribution will be determined whenever a significant core configuration change is made. 5.0 DESIGN FEATURES Those design features relevant to operation safety and to limits that have been previously specified are described below. These features shall not be changed without appropriate review. 5.1 Reactor Fuel Fuel elements shall be of the general MTR/0RR type with thin plates containing uranium fuel enriched to about 93% 235 u and clad with aluminum. The fuel matrix may be fabricated by alloy-ing high purity aluminum-uranium or by the powder metallurgy method where the starting ingredients (uranium-a luminum) are in fine powder form. Fuel matrix may also be fabricated f rom uranium oxide-aluminum (U 0 -A1) using the powder metallurgy 3g process. Elements shall conform to these nominal specifications: Overall Size: 3 in. x 3 in. x 34 in. Clad Thickness: 0.015 in. 275 171

35. Plate Thickness: 0.050 in. Water Channel Width: 0.12 in. No. of Plates: standard element - 16 fueled plates (min.) window element - 16 fueled plates (min.) control element - 9 fueled plates (min.) partial element - 9 fueled plates (min.) Plate

Attachment:

swaged or pinned 235 Fuel Content (Total): 200 g U nominal. Fuel Burnup: The fuel burnup shall not exceed.94 x 10 3 fission /cm except for U O -Al which shall 3g 21 3 not exceed 1.5 x 10 fission /cm. 5.2 Control and Safety Systems Design features of the components of this system (3.2.2, 3.2.3) that are important to safety are given below. 5.2.1 Power Level (Normal Channels) For this function three independent measuring channels are p ovided, two of which are required to be operable as a minimum. Each channel covers reliably the range from about 25% to 150% (of 5 MW). Each channel comprises an uncompensated baron-coated ion chamber feed-ing an ampli fier that controls electronic swi tches in the DC cur-rent that flows through each control rod electromagnet. Each chan-nel controls and scraras all control rods. Each channel is fail-safe. The "f as t" scram (% 5 ms) from each channel also produces, and is backeJ up by, a " slow" scram (% 2G ms) through interruption of the AC supply to the rod electromagnet DC power supply. Each channel indicates power level on a panel metet allowing channel checks to be done during reactor operation. Each chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power. 275 172

36. 5.2.2 Power Level (In termediate) Channel For this function a single channel is provided, covering reliably the range 10~ % to 300% (of 5 MW) with a logarithmic output indi-cation on both a panel meter and a chart recorder. To cover the range under all core condi tions a gamma-compensated boron-ion cham-ber is used to supply a logarithmic amplifier. The chamber can be changed in position, over a limited range, so as to allow the chan-nel reading to be standardized against reactor thermal power. Rate of change of power information is also derived, in the form of a per-iod, that can produce a fast scram (and backup slow scram) in the same way as in Sect. 5.2.1. From this channel is also derived control and inhibit actions, viz. bypassing ofcount rate channel functions, bypassing of flow and flapper scrams, reversal and in-hibit of control rods. To negate the ef fect of overcompensation in the ion chamber, which can occur under certain conditions even in an initially undercompensated chamber, provision is made to supply an adjustable small current to the channel amplifier (up -10 to'l.5 x 10 A) so as to facilitate startup. 5.2.3 Count Rate channel A fission chamber is used to supply pulses to a linear amplifier and logari thnic count rate ci rcui try. Pulse height discrimination selects pulse ampli tudes that correspond to fission events and rejects those from alpha particles. Count rate on a logari thmic scale is displayed on a panel meter and a chart recorder. In addition, count rate period information is derived and similarly displayed. The channel covers a range of 1-10 cps, correspond-ing roughly to 0.25 mw - 2.5 L', but the upper limi t can be in-creased many decades by reposi tioning the chamber. The motor-operated chanber drive is operated from the control room, the drive posi tion being indicated on a meter. To prevent control-rod wi thdrawal when the neutron count rate information may not be reliably indicated, inhibits are provided on counc rate and period, and when the fission chamber is being reposi tioned. All 275 173

37 these inhibits are bypassed at a power of > 50 W. A scaler is also provided for obtaining accurate values at low count rates if needed (e.g., approach to cri tical wi th new fuel or new core con fi g rua t i on ). 5.2.4 Neutron Source For obtaining the reliable neutron information necessary for startup from a cold sh'at-down condition, an antimony-beryllium neutron source is provided for insertion into the core as needed. 124 This source, nominally 50 curies of Sb, is renewed by neutron activation in the core. Its presence in the core is not essen-tial exceot af ter extended shut-downs. Integrity of the source is checked by periodic sampling of pool water (4.9). 5.3 Rod Control System e J.1 Control Rods Up to five control rods are provided for the control of core re-activity. These rods may be either boron-carbide or silver-in-di um-cadmium (see 4. 3.1). Individual integral worths vary f rom about 1-4% a K, depending on position and core configuration. The rods are coupled to drive shafts through electromagnets that allow release of the rods within 50 ms af ter receiving a scram signal. Position indicators on the control console show the ex-tent of withdrawal for each rod and a digital readout can be switched to any one rod. To limit the. ate of reactivi ty increase upon startup, the rod drive speeds are limi ted to 5 in./mb. and no more than two rods can be withdrawn simultaneously. Switches on the guide tubes attached to the control fuel elements are arranged to produce a scram if any guide tube is lifted. This guards against lif ting of the attached fuel element. 5.3.2 Regulating Rod One regulating rod is provided to aid in fine contro, and main-tenance of constant reactor power for long periods. The rod is non-fueled, is limi ted to a total worth of 0.6% a K fo,r safety E ]fk

38. reasons (3.1.4), and can be ei ther manually or servo-controlled. The drive speed is 24 jaAnh. Coarse and fine position readouts are provided. 5.4 Cooling system 5.4.1 Primary Cooling System Core cooling is ef fected by gravity flow of demineralized water from the reactor pool to an underground holdup tank that provides an approximate 10 minute delay to allow ' N activi ty to decay. The water is then pumped back to the pool through the primary side of a heat exchanger where heat is transferred to a secondary cool-ing system. The holdup tank is vented to the building exhaust duct. The driving force for the coolant is the fixed head between the pool overflow gutter and the water level in the holdup tank, the latter being fixed by the total volume of water in the system. Flow is adjusted to a desired amount with a valve in the core exit line. Core cooling is not immediately affected by pump failure as flow will continue till the water levels equalize; neither will the pool be drained. To prevent leakage of water through the pool walls, a continuous steel shell is located within the concrete pour of the pools. All embedments penetrating the pools are welded to this shell. To eliminate corrosion of inaccessible piping, the embedded portion of the reactor primary cooling piping under the pools is stainless steel. To change over automatically to natural convection cooling at low flow rates, a weighted flapper valve seals the core exit plenum. This valve, held closed by the core pressure drop, opens by gravi ty when the flow drops to approxi-mately 700 gpm. Leakage at the flapper valve seat, or in the plenum, is monitored by a plenum leak detector that senses planum pressure increase and alarms in.he centrol room. Primary flow is neasured by taking the pressure drop across an ori fice plate in the core exit line, indication be'ng both in the pump room and on a recorder in the control roon, Temperature sensors in the pool (above the core) and in the core exi t line allow the 275 175

39 core A T to be measured. These are resistance thermometers having alternative recorder and digital readout in the control room. Float switches are provided to moni tor pool level. Normal pool level (a t the overflow gutters) provides 24 f t. of water over the top of the fuel. 5.4.2 secondary cooling system Reactor power transferred through the heat exchanger is dissipated to the atmosphere via a cooling tower. To minimize corrosion, the exchanger has stainless steel shell and tubes. To prevent water from entering the secondary system should a tube leak occur, the static pressure in the secondary is made higher than that of the primary through the relative elevations of the two systems. 5.4.3 core soray For backup in the event a hot core is exposed, two spray nozzles are located at the two alternative operating positions of the re-actor. They are controlled from a manually-operated valve located outside the reactor building. 5.5 containment system 5.5.1 Physical Features The containment structure consists of the reactor building, with 3 a free ai r volume of about 7700m. This building houses the re-actor, the primary cooling system including holdup tank, and the heat exchanger. Personnel access is via double ai rlock doors or sliding doors with inflatable seals. A 4 2-feet deep water-filled canal penetrates the building with containment provided by a 25-inch deep water-seal weir. A water-tight gate wi th inflatable seals can be used to shut the canal of f f rom the reactor pools when needed. Ventilation access to the building is through pneumatically operated damper valves that can be used to seal the building. These dampers are fail-shut upon reduction of ai r pressure. 275 176

40. 5 5.2 Emergency Secuence While negative pressure within the containment building is not a requisite for reactor operation, it is required in the event of a release for the controlled-release containment of ai rborne radio-acti ve material. The emergency sequence is initiated either auto-matically by the excursion monitor (see 3.3) or manually by the console operator. The sequence is that all air supply ducts and the pool sweep dampers are closed immediately, followed later by the exhaust duct damper as soon as negative pressure (-1 in. w.g.) is attained in the building (but not more than 7 sec. later). Closure of the pool sweep damper prevents activity released above the core from reaching the exhaust duct before it closes. Also closed immediately are the isolation valves in the vent line and the air purge to the holdup tank. This prevents activity in this tank from reaching the exhaust duct. Upon closure of the exhaust duct damper, the emergency exhaust fan starts and maintains a nominal negative pressure in the containment building. This fan exhausts building air at a low rate (<200 cfm) through absolute and charcoal filters before connecting into the nornal exhaust duct. The latter discharges to the atmosphere through a stack at a high elevation. The entire evacuation sequence is fail-safe upon loss of utility electric power. It will operate wi th ei ther utility or emergency generator power. 5.5.3 Exhaust Duct Moniter (" Stack Moni tor") Ai r in the exhaust duct is continuously sampled for carticulate, iodine, and gaseous activities each being read by separate de-tectors. The relative proportions of each type of activity can thus be dete rmi ned. The results are indicated on chart records, with repeaters in the control room. Detection or indication of a release is not dependent on all three detectors bei..g opera-tional, for any release will have associated wi th it all three types c' activity or will af fect each detector to sone extent. Alarms wh n set points are exceeded are given at the monitor and repeated in the con trol room. 275 177 -.. _. ~ _ -

41. 5.6 Fuel Storage 5.6.1 New Fuel Uni rradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Plan. Elements are stored upright in metal racks in which the separation between elements is a minimum of 2 inches. With such an arrangement, subcriticality is assured ( if. 5). 5.6.2 Irradiated Fuel irradiated fuel is stored upright under water in the storage pool within the reactor building in criticality-safe racks. Each reck accommodates 16 elements in wells with center-to-center spacing of 6 inches. Ref. 5 states that an infini te number of elements so stored would be subcritical. Each well has a bottom hole to allow circulation of water for cooling. 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The organiza; ion for the management and operation of the reactor facility shall be as a minimum the structure shown in Fig. 2. Job titles shown are for illustration and may vary. Four levels of authori ty are provided, as folicws: Level 1: Individual responsible for the facility license and site administration. Level 2: Individual responsible for the reactor facili ty opera-t i on and managemen t. Level 3: Individual responsible for daily reactor operations. Level 4: Reactor ope rating staf f. The Nuc aar Safeguards Committee shall report to Level 1. Radia-tion safety personnel shall report to Level 2 or higher. 275 178

42. 6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in Figure 2. Management levels in addition to having responsibility for the policies and operation of the reactor faciiity shall be responsible for safeguarding the public and facili ty personnel f rom undue radiation exposures and for adhering to all requirements of the operating license and tech-nical speci fications. In all instances responsibilities of one level may be assumed by designated alternates or by higher levels, condi tional upon appreoriate quali fications. 275 179

43. FIGURE 2 ORGANIZATIONAL STRUCTURE Vice-Presicent Nuclear Products LEVEL 1 I Manager, Nucleonics l Operating Manager l Sterling Forect Labi I i Nuclear Safeguards 4 Cc=mittee l 1 l i l l l 7! Manager, Radiochemical Manager, Heal *.h, Saf ety,; l Production l& Environnental Affairs _ LEVEL 2 Manager, Nuclear Operations i J l l LEVEL 3 Reactor Supervisor Reactor Operations Staff 1 l 2-Chief Operator i i LEVEL 4 l Lead Reactor Operators i Reactor Operators 1 I (LicensedOperatorsat.dTrainees) N 275 100

44. 6.1.3 Staffing a. The minimum staf fing when the reactor is not secured shall be: (1) A licensed Reactor Operator in the control room. (2) A second licensed reactor operator Dresent at the reactor facility. Unexpected absence for two hours is acceptable provided immediate action is taken to obtain a replacement. (3) A licensed Senior Reactor Operator shall be readily available on call. (4) A member of the operating shift shall be designated by Level 2 management as knowledgeable in radiation control. b. Events requiring the presence of a Senior Operator: (1) All fuel element or control-rod alterations within the reactor core region. (2) Relocations of any experiments with reactivity worth t 1 dollar. (3) Recovery from unplanned or unscheduled shutdowns unless they are of a type excluded by the Level 2 authority. Such exclusions shall be posted in the control room or placed in the appropriate procedures. Furthermore, the presence of a senior operator at the facility shall not be reauired during recovery from unplanned or unscheduled shutdown or significant re-duction in power in instances which result from: 1. Electrical power interruptions from internal or external failures exclusive of power supply failures of the reactor instrumentation, control and safety systems; 2. False signals, which, in the opinion of the Senior Operator, were properly verified to be false and to have resulted from monitoring, experimental, or cr-trol equipment, or from personnel inadvertence; and 3. Inten donal shutdowns made L. le Reactor Operator which are not related to the safety of she reactor; provided that prior to the initiation of such recovery, the Senior Operator shall be notified of the shutdown of pcwer reduction, and sha!1 determine that the shutdown was caused by one of the enumerated occurrences, and shall determine that his presence at the facility during recovery is not required E h0l

45. 6.1.4 Selection and Training of Personnel The selection, training, and requalification of personnel shall meet or exceed the requirements of ANS-15.4/N 380 and Appendix A of CFR Part 55 and be in accordance with the requali fication plan approved by the Commission. 6.1.5 Review and Audit The independent review and audit of reactor facility operations shall be performed by the Nuclear Safeguards Committee. 6.1.5.1 Composi tion and Quali fications The Nuclear Safeguards Committee'shall be composed of a minimum of 5 members. The members shall collectivelv provide a hroad spectrum of expertise in the appropriate reactor technology. Mem-bers and alternates shall be appointed by and report to the Level 1 authority. They may include individuals f rom within and/or out-side the operating organization. Qualified and approved alter-nates may serve in the absence of regular members. 6.1.5.2 Charter and Rules The committee shall function under the following operating rules: a. Meetings shall be held not less than Semi-annually or more f requently as ci rcumstances warrant consistent with effective monitoring of facility activities, b. A quorum shall consist of not less than one half the member-ship, where the operating staff does not constitute a ma-jori ty. Sub groups may be appointed to review specific items. c. d. Minutes shall be kept, and shall be disseminated to members and to the Level 1 iuthori ty wi thin ene month af ter the meeting. fh

46. The Committee shall appoint one or more qualified individuals e. to perform the Audit Function. 6.1.5.3 _ Review Function The following items shall be reviewed by the review group or a subgroup thereof: Determinations that proposed changes in equipment, systems, a. tests, experiments, or procedures do not involve an unre-viewed safety question. b. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety signi ficance. Tests and experi ents in accordance with section 6.3 c. d. Proposed changes in technical specifications, license, or

charter, Violations of technical specifications, license, or charter.

e. Violations of internal procedures or instructions having safety significance. f. Operating abnormalities having safety significance, and audit reports. g. Reportable occurrences listed in section 6.5.3 6.1.5.4 Audit Function The audi t function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Where necessary, discussions with responsible personnel shall take place. In no case snall the individual or individuals conducting the audit be immediately responsible for the area being audited. The following items shall be audited: a'. The conformance of facility operations to the technical specifications and applicable license or charter condi-tions, at least once per calendar year (in terval not to exceed 18 months). f h][

47. b. The retraining and requrlification for the operating staff, at least once every other calendar year (interval not to exceed 30 months). c. The results of actions taken to correct deficiencies occur-ring in reactor facility equipment, systems, s t ructures, or methods of operations that affect reactor safety, at least once per calendar year (interval not to exceed 18 months). 4w.. d. The reactsr facility Security Plan and imple-menting procedures at least once every other calendar year (interval not to exceed 30 months). Deficiencies uncovered that af fect reactor safety shall immediately be reported to the Level 2 authority. A written report of the findings of the audit shall be submitted to the Level 1 authori ty and the Nuclear Safeguards Commi ttee memoers within 90 days after the audit has been completed. 6.2 Procedures There shall be wri tten procedures for, and prior to, ini tiating any of the activities listed in this secticn. The nrocedures shall be reviewed by the Nuclear Safeguards Committee and approved by Level 2 or designated alternates, and such reviews and approvals shall be documented. Several of the following activities mav ha included in a single manual or set of crocedures nr div. dad emenq various manuals or procedures. a. S t a r t uo, ope ra t i on, and shutdown of the reactor, b. Fuel leading, unicading, and movement within the reactor. c. Routine maintenance of major components of systems that could have an effect on reactor safety. d. Surveillance tests and calibratiens required by the technical scecifications or those that may have an ef fect en reactor safety.

43. Personnel radiation protection, consistent wi th applicable e. regulations. f. Administra, controls for operations and maintenance and for the conv...; of irradiations and experiments that could affect reactor safety or core reactivity. g. Implementation of the Security Plan. Substantive changes to the above procedures shall be made only .af ter documented review by the Nuclear Safeguards Commi ttee and ' approval by Level 2 or designated al ternates. Minor modi fica-tions to the original procedures which do not change their ori-ginal intent may be made by the Level 3 authority (Reactor Supervisor) and must be approved by Level 2 or designated al-i ternates within 14 days. Temporary changes to the procedures that do not affs.ct reactor safety may be made by a Senior Re-actor Operacor and are valid for a period of one month. Such temo-orary changes shall be documented and reported to Level 2 or designated alternate. 6.3 Excerirent Review snd Acoroval a. All new experiments or classes of experiments that could affect reactivity or result in release of radioactive materials shall be reviewed by the Nuclear Safeguards Con-mittee. This review shall assure that compliance with the requirements of the license, techni cal speci fications, and applicable ragulations has !eer satisfied, and shall be documented. b. Pri or te review, an experiment plan or precosal shall be prepared describing the experiment including any safety con s i de ra t i on s. Review commen ts of the Nuclear Safeguards Commi ttee setting c. forth any conditiens and/or limitations shall be documented in Comc' ttee minutes and submi tted to Level 2. 275 185

49 d. All new experiments or classes of experiments shall be approved in writing by Level 2 or designated alternates prior to their initiation. e. Substantive changes to approved experiments shall be made only af ter review by the Nuclear Safeguards Commi ttee and written approval by Level 2 or designated alternates. Minor changes that do not significantly alter the experiment may be approved by the Level 3 authori ty (Reactor Supervisor). f. Approved experiments shall be carried out in accordance with established approved procedures. 6.4 Recuired Actions 6.4.1 Action to be Taken in Case of Safety Limi t Violati on a. The reactor shall be shutdown, and reactor operations shall not be resumed until authorized by the Commissicn. b. The safety limit violation shall promptly be reported to the Level 1 authority or designated alternates. c. The safety limit violation shall be reported to the Com-mission in accordance with section 6.5 3 d. A safety limi t violation report shall be prepared. The report shall describe the following: '1) Aoplicable ci rcumstances leadi,g to the violation. (2) Ef fect of the violation upon reactor facility com-ponents, systems, or structures. (3) Corrective acticn to be taken to prevent recurrence. The report shall be reviewed by the Nuclear Safeguards Cceni ttee. A follow up report describing extant ac*ivities shall be sub-mi tted to the Ccmmi ss icn when authori zation is scught to resume operaticn of the reactor. 275 186

50. 6.4.2 Action to be taken in the event of an occurrence as defined in section 6.5.3, a-1, 3: a. Corrective action shall be taken to return conditions to normal; otherwise, the reactor shall be shut down and reactor operation shall not be resumed unless authorized by the Level 2 authority or designated alternates. b. All such occurrences shall be promptly reported to the Level 2 authori ty or designated alternates. c. A!! such occurrences where applicable sFall be reported to the Commi s s i on in accordance with section 6.5.3 d. All such occurrences including action taken to prevent or reduce the probability of a recurrence shall be reviewed by the Nuclear Safeguards Commi ttee. 6.5 Reoorts in addi tion to the requi rements of applicable regulations, reports shall be made to the Commissicn as follows: 6.5.1 Startuo Re= orts Three months af ter completion of requisite s tartup and power-escalatien testing of the reactor, or nine mcnths af ter criti-cality, a written report shall be submitted to the Ccmmission. The report shall include a summary of the following: a. Description of measured values of operating conditions or characteristics obtained and comparison of these values with design predictions or specifications. b. Descriptiens of major corrective actions taken to cbtain sati s f actory operation. c. Re-evaluatien cf safety analyses where measured values in-dicate substantial variance f rom those values use. in the Safety Analys i s Report. 275 187

51. 6.5.2 Coerating Recorts Routine annual reports covering the activi ties of the reactor facili ty during the previous calendar year shall be submitted to the appropriate NRC Regional Office with a copy to tne Director of Inspection & Enforcement within 3 months following the end of each prescribed year. Each annual operating report shall include the following information: A narrative summary of reactor operating ex serience including a. the energy produced by the reactor. b. The unscheduled shutdowns including, where applicable, cor-rective action taken to preclude recurrence, but excluding those of the types listed in Sect. 6.1.3. b (2) above. c. Tabulatien of major preventive and corrective maintenance operations having safety signi ficance. d. Tabulaticn of major changes in the reactor facility pro-cedures, and new tests and/or experiments signi ficantly different from those performed previously anc which are not described in the Safety Analysis Report, including cenclusions that no unreviewed safety questions were involved. A summary of the nature and amount of radioactive ei'luents e. free the reactor facility released or discharged to the environs. Th e s umma ry s ha l l include where practicable an estimate of individual radionuclides present in the ef fluent if the es-timated average release af ter dilutien or diffusion is greater than 25% of the cencentration allcwed er reccmmended. f. A sur ary of exposures received by f acili ty personnel and visi tors.vhere sucn exposures are greater than 25% of that allowed or recemrenced. g. A summary of tne calculated doses to a critical individual in the unrestricteo area due to the airoorne releases of noble gases and raciciodines. 27'c 188

52. 6.5.3 Soecial Reports (Reportable Cccurrences) a. There shall be a recort not later than the following working day by telephone and confirmed by telegraph or similar con-veyance to the Commission to be followed by a wri tten report within 14 days of any of the folicwing: (1) Release of radioactivity from the reactor above allowed limits, as provided by section 3.8.1 of this specifica-tion. (2) Violation of Safety Limi ts (3) Any of the following: (a) Operation with actual safety-system settirgs less conservative than the limi ting safety-system set-tings specified in the Technical Speci fications. (b) Opera ti on in violation of Limi ting Conditiens for Operation established in the Technical Specifica-t i on s. (c) A reactor safety system condonent malfunction which renders or could render the reactor safety system incapable of performing i ts intended safety functicn unless the malfunction or conditicn is discovered during tests or periods of reactor shutdowns. (Note: Where components or systems are provided in addition to those requi red by the Techni cal Speci-fications, the failure of the extra components or systems is not considered reportable provided that the minimum number of conocnents or systems speci-fled or required perform their intended reactor safety function.) (d) An unanticicated or uncontrolled change in reactivity greater than or equal to 1% 2K/K. (e) Abnormal and signi ficant degradation in reactor fuel, and/or cladding, coolant bcundary, or con tainmen t I

4 53 boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radia-tion exposure limits of personnel and/or envi ron-ment. (f) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused an ur.- safe condition with regard to reactor cperations. b. A written "eport wi thin 30 days to the Commission of: (1) Permanent changes in the facili ty organization structure. (2) Significant changes in the transient or accident analysis as described in the Safety Analysis Report. 6.6 Records Records of the following activities shall be maintained and re-tained for the periods specitled below. The records may be in the form of logs, data sheets, or other suitable forms. The re-quired information may be contained in single, or multiple records, or a coebination thereof. Recorder charts showino nceratino nara-meters of the reactor (i.e., power level, flow, temoerature, etc.) for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum ceriod of rso years. 6.6.1 Records to be Retained for a Period of at least Five Years or for the Life of the Component involved whichsver is smaller. a. Normal reactor facili ty operations (including scheduled and unscheduled shutdowns). Note: Supporting documents such as checklists, log sheets, etc. shall be maintained for a period of at least two yea rs. b. Principa l maintenance ope ra t ic,s. c. Reportacle occurrences. d. Surveillance activi ties recui red by the Technical Speci fi ca t ion s. k)0

1 i. 54. e. Reacter facili ty radiation and contamination surveys where rer,uired by applicable regulations. f. Experiments performed wi th the reactor. g. Special Nuclear Materials (SNM) inventories, receipts, and s h i pmen ts. h. Approvec changes in operating procedures,

i. Records of meeting and audit reports of the Nuclear Safe-guards Committee.

J. Sealed Source leak test results. 6.6.2 Records to be Retained for at least One Requalification Cycle or for the Length of Employment of the Individual whichever is Smaller: a. Retraining and requalification of licensed operations per-sonnel. However, records of the most recent complete cycle shall be maintained at all times the individual is employed. 6.6.3 Records to be Retained for the Li fetime of the Reactor Facility: (Note: Annual reports may be used where applicable as records in this section.) a. Gaseous and liquid radioactive effluents released to the

environs, b.

Of f-s i te envi ronmental-moni toring surveys requi red by the Techni cal Speci ficati ons. c. Radiation exposure for all personnel moni tored. d. Updated drawings of the reactor facility. 275 191

n 55.

7.0 REFERENCES

1. Supplement No. 2 to Final Hazard Summary Report (April 1977). 2. Final Hazards Summary Report (May "J60). 3 Reactor Power Excursion Tests in the SPERT IV Facility, 100-17000 (Aug. 1964). 4. Supplementary Informatien to Final Hazards Summary Report (28 Apr. 1961). 5 Cri tical Experiments with SPERT-0 Fuel Elements, ORNL-TM-1207 (July 14,1965), by E. B. Johnson and P. K. Reedy, Jr. 6. Supplement No. 3 to Final Hazard Summary Report (Dec. 1977). 7 Supplenent No. 4 to Final Hazard Summary Report ( Ao r. 1978), 275 192 .}}