ML16273A304

Us Department of the Interior, Geological Survey, Responses to E-Mail Questions

ML16273A304
Person / Time
Site:	U.S. Geological Survey
Issue date:	09/22/2016
From:	Roy B R US Dept of Interior, Geological Survey (USGS)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16273A304 (44)

Text

{{#Wiki_filter:llUSGS science for a changing world Department of the Interior US Geological Survey PO Box 25046 MS 974 Denver, CO 80225-0046 September 22, 2016 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Reference:

U.S. Geological Survey TRIGA Reactor (GSTR), Docket 50-274, License R-113, Email from Mr. Wertz dated September 20, 2016

Subject:

Responses to email questions Mr. Wertz: Responses to your email on September 20, 2016, are provided in the enclosed pages. Please contact me if further details, or corrections, are needed. As per our phone conversations during September 2016, I have attached a "track changes" version of the proposed Technical Specifications as Attachment

1. Sincerely, Brycen R. Roy USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct. Executed on 9/22/2016 Attachment Copy to: Vito Nuccio, Reactor Administrator, MS 911 USGS Reactor Operations Committee

'\ USGS response to USGS Technical Specification

1. Propose to revise TS 4.1, Specification 3, to read, (Response to RAI No. 17.b) "3. The maximum reactivity insertion rate of a standard control rod shall be measured annuafly or following a change in core or control rod configuration that is expected to change the total reactivity worth of that control rod by more than $0.30 (not including transient fission product poison effects)." Response:

We agree with the above recommendation.

2. (Response to RAI No. 19.a) --Currently, TS 3.3 to read: "c. The reactor shall not be operated if the tank water level is more than 24 inches below the top lip of the reactor tank, and an alarm which is audible to the reactor operator shall sound when the water level is too low." Proposed revision TS 3.3 to read: "c. The reactor shall not be operated and an alarm which is audible to the reactor operator shall sound if the tank water level is more than 24 inches below the top lip of the reactortank." Response:

We agree with the above recommendation.

3. TS 5.4 states: 1. All fuel elements and fueled devices shall be stored in a geometrical array where the effective is less than 0.9 for all conditions of moderation.

ANSI/ANS has "and reflection." Proposed revision TS 5.4, Specification

1. to read: 1. All fuel elements and fueled devices shall be stored in a geometrical array where the effective is less than 0.9 for all conditions of moderation and reflection.

Response: We agree with the above recommendation.

4. Response to RAI No. 33 TS 6.1.2 does not have the following sentence as stated in the RAI response.

Proposed to add the following sentences to the beginning of TS 6.1.2. "Responsibility for the safe operation of the reactor facility shall be with the chain of command established in Figure 1. Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license, the established charter, and the technical specifications. Response: This was an oversight in making the TS edits. We agree with the above recommendation and will add the sentence as stated in the previous RAI response.

5. TS 1.2 Definitions, "Control Rods,":

• Delete an extra space between section and Its. 2. Shim Rod: A shim rod is a control rod having an electric motor drive and scram capabilities.

It may have a fueled-follower section. Its position is varied manually.

• Delete an extra space between capabilities and It. 3. Transient Rod: The transient rod is a control rod having an electric motor and pneumatic cylinder drive with scram capabilities.

It can be rapidly ejected from the reactor core to produce a pulse or its position may be varied manually. It may have an air-filled follower. Response: We agree with the above recommendation.

6. TS 3.1.3, Specification 2 Replace commas with a period after 2. Response:

We agree with the above recommendation.

7. TS 3.2.3, Specification 1 Replace commas with a period after 1. Response:

We agree with the above recommendation.

8. TS 5.4, Specification 3 Remove an extra space between to and TS 3.3, Specification 1.b 3. If stored in water, the water quality shall be maintained according to TS 3.3, Specification 1.b. Response:

We agree with the above recommendation.

9. TS 6.1.3, Specification 2 Place a colon after Specification 2 to read "2. Events requiring the direction of a Senior Reactor Operator:" Response:

We agree with the above recommendation.

10. TS 6.1.3, Specification lb Remove a space between second and person b. A second person present within the Denver Federal Center who is able to carry out prescribed instructions; J*,* i :i. Response:

We agree with the above recommendation.

11. TS 6.1.4 Selection ofTraining of Personnel Remove an extra space between 15.4 and "Selection

... " The selection, training and requalification of operations personnel shall follow the guidance of ANSI/ANS 15.4, "Selection and Training of Personnel for Research Reactors." Response: We agree with the above recommendation.

12. Title TS 6.2.4, Audit Function -needs to be Bold and consistent font size. Response:

We agree with the above recommendation.

13. There seems to be a comment on page 33, in TS 6.3 Response:

We will remove the comment reference above. 14. TS 6.6.2, Specification 4 Remove an extra space between TS and 6.7.2 4. Where appropriate, a report shall be submitted to the NRC in accordance with TS 6.7.2. Response: We did not find the referenced extra space, so we believe no change is necessary.

15. TS 1.0 Definition of "Reference Core Condition," need to change "negligible" to "zero" due to SDM of $0.30. ANSI/ ANS-15.1-2007 has "negligible

(<0.30 dollar)" which if used could negate TS limit. Response: As per our phone conversation on 9/21/16, we will change the definition to read as follows: Reference core condition: The condition of the core when it is at ambient temperature (cold, 18-25 0 C) and the reactivity worth of 135 Xe is less than $0.01. 16. TS 3.5, Specification 4, needs "the" before "ventilation" in the last sentence. Response: We agree with the above recommendation.

17. TS 4.3, Specification 3, has an extra space between the two sentences.

Response: We agree with the above recommendation.

18. TS 5.1, Specification 3, needs "a" before "vertical." Response:

We agree with the above recommendation.

19. TS 6.1.3, Specification 1.d, remove 2nd instance of "on-call" as it would allow operation with only one person (SRO) at the facility.

Response: We propose to remove the proposed Specification l.d entirely. With Specification l.d removed, and in the condition where the Licensed Operator in the control room is a SRO, Specification 1.a would still be met. Specification l.b would still have to be met, but Specification l.c for having a SRO on call would not be required (as stated in the specification "If neither of these two individuals is a Senior Reactor Operator"). As a part of this change, Specification l.e would be changed to 1.d 20. TS 6.2.1, lots of "double spaces." Response: We agree with the above recommendation.

21. TS 6.2.1, this sentence: "To expedite Committee business, a Committee Chairman may be appointed" seems to contradict the first sentence which requires a Committee Chairman.

Consider deleting sentence. Response: We agree with the above recommendation and will reword the first paragraph of TS 6.2.1 as follows: The ROC shall be composed of at least four voting members, including the Chairman. All members of the Committee shall be knowledgeable in subject matter related to reactor operations. To expedite Committee business, a Committee Chairman shall be appointed.

22. TS 6.2.2, first sentence, change "must" to "shall." Response:

We agree with the above recommendation.

23. TS 6.2.3, Specification 1, change to read: "Determinations that proposed changes in the facility and procedures, and the conduct of tests or experiments are allowed without prior authorization by the U.S. NRC, as detailed in 10 CFR 50.59;" to align with requirements in 10CFR50.59.

Response: We agree with the above recommendation except we will use the term NRC instead of U.S. NRC as defined in our initial use of U.S. Nuclear Regulatory Commission.

24. TS 6.2.4, Specification 1, remove "or charter' since it only applies to the ROC. Response:

We agree with the above recommendation.

25. TS 6.6.1, Specification 2, add "and" before ROC. Response:

We agree with the above recommendation.

26. TS 6.7.2, Specification 1, add "confirmed by" before "digital" to match ANSl/ANS-15.1.

Response: As per our phone conversation on 9/21/2016, we propose the following wording to TS 6.7.2, Specification 1: A report within 24 hours by telephone, confirmed by digital submission or fax to the NRC Operations Center if requested, and followed by a report in writing to the NRC, Document Control Desk, Washington, D.C. within 14 days that describes the circumstances associated with any of the following:

27. TS 6.7.2, Specification e, revise to match ANSl/ANS-15.1: "Malfunction of a required reactor safety system component which renders or could render the system incapable of performing its intended safety function unless the malfunction or condition fa+ltH:e is caused by discovered during maintenance, then no report is required tests or periods of reactor shutdown;" Response:

We agree with the above recommendation.

28. TS 6.7.2, Specification g, revise to match ANSl/ANS-15.1: "An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operationscondition which results or could result in operation of the reactor outside the specified safety limits; or" Response:

We believe no change to the wording is required. We feel like the statement "unsafe condition with regard to reactor operations" is ambiguous and opens the door to the licensee and regulator having different interpretations. Use of the phrase "condition which results or could result in operation of the reactor outside the specified safety limits" is better because reactor safety limits are clearly defined throughout the Technical Specifications. APPENDIX A To FACILITY LICENSE NO. R-113 DOCKET NO. 50-274 TECHNICAL SPECIFICATIONS AND BASES FOR THE UNITED STATES GEOLOGICAL SURVEY TRI GA RESEARCH REACTOR SEPTEMBER 2016 , '"' " "' ',* ,, , ,>, '.- TECHNICAL SPECIFICATIONS AND BASES FOR THE US.GS TRIGA ,R,E,SEARCH,REACT?R. , ", .. 1 '.r,, 1. Introduction

• .*' I*' *,'t 1.1 Scope '*.,, This document constitutes the Technical Specifii:ationdor the Fcicilityllcense 113 as requfi-ed*b*y ib CFR 50.36 and supersedes all prior Technical SpecificatiOns.

This'document includes the "Basis" to support the selection and significance of the specifications. Each basis is included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere, except whe're they reference the USGS SAR or a* specific Technical Specification. These specifications are formatted in a manner consistent.with ANSI/ANS 15.1-2007. . ; I *,*I* 1.2 Definitions Audit: A quantitative examination of records, procedures or other documents. . . )' * ' ' * , '. ., v_. Channel: A channel is the combination of sensing, signal processing, *and o'utputting'devices which are connected for the purpose of measuring the value of a *parameter.'*

• * " ' .' . Channel Calibration:

A channel calibrc:i.t.iofl an adjustment of the_ channel_ such that its output.. corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire equipment actuation, alarm, or trip and shall include a Channel Test. :,1* Channel Check: A channel check is a qualitative of acceptable'performance by of channel behavior. This verification, whe.re possible, shall include comparison 'of the channel with other independent channels or systems measuring the same variable. Channel Test: A channel test is the introduction of a signal into the:chinriel for verification that it is operable. Confinement: Confinement means an enclosure of the reactor bay which is designed to limit the release of effluents from the enclosure to the external environment th rough' controlled of defined pathways. Control Rod: A control rod is a device fabricated from borated graphite, B 4 C powder or boron and/or fuel which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unifallciwirig it to perform a' safety function whe'n the coupling is disengaged. Types of control rods shall include: 1. Regulating Rod (Reg Rod): The regulating rod is a control rod h'aving e1ectric motor drive and scram capabilities. It may have a fueled-follower section. Its position may be varied manually or by the servo-controller. 1 'I i *, . ..

• 2. Shim Rod: A shim rod is a control rod having an driJe and it may have a fueled-follower section. -Its position is varied manually.

3. Transient Rod: The transient rod is a control rod having an electric motor and pneumatic cylinder drive with scram capabilities. -It can be rapidly ejected from the reactor core to produce a pulse or its position may be varied manually.

It may have an air-filled follower. Excess Reactivity: Excess reactivity is that amo.unt of reactivity th?t 1,11,oylo all c,:ontrol rods were moved to the maximum reactive condition from the point where the.reactor is exactly critical (k.tt:;=l) at reference core conditions. Experiment: Any operation, hardware, or target (exclµding r;i_evices-such which is,designeci to investigate non-routine reactor characteristics or,which*is intended for irradiation within an* irradiation facility. Hardware rigidly secured to a core or shield structure so as to be a part of their design to carry out experiments is not normally considered an experiment. Specific experiments shall include: 1. Secured Experiment: A secured experiment is any or exp'eriment' . that is held in a stationary positipn relative to the r.eactor by mechanical means. Jhe restraining forces must be substantially greater than those to which the exper,im.ent .,,; *.. subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces that can *arise as a result of ,:.: * * ' I *

• I ,

2. Movable Experiment:

A movabl.e experiment is one that. not secpre.d and intendeditp be,: ... moved while near or inside the core during reactor operation. Instrumented Fuel Element: An instrumented fuel element is .a special fuel e\ement in orie.or;,* (: 2; more thermocouples have been embeqded for the purpoSIJ:of.measuring the fue,l terl)peratu.res .ciuril),!l, " . reactor operation. Irradiation Facilities: Irradiation facilities shall. 111ean vertica_l. tubes, rotating, spedmen rac;k, pneumatic transfer system irradiation tubes, sample-holding dummy fuel elements and any other in-tank device intended to hold an experiment. '* I j t '* * . ,; '* Licensed Area: Rooms 149-152, 154, 157, 158, 610, BlOB and o,f,Bupqing the .area insipe.the wrought iron fence and south cooling tower wall that is near the SW corner of Building 15; and Room 2 of Building 10. ,. * *. * *i *1* ' _' )' "' Measured Value: The measured value _is, the ,l'alu,e, cm the output of a channel. Operable: A system or component shall qe con,siqereq. j_t is of performing its intended function. 1,.:. ,* ,. ':*. !1 ,-Operating: Operating means a component or system is performing its intended function. 2 *)-* > * , ' ' *, . , . ' Pulse Mode: Pulse mode shall mean any operation of the reactor with the mode selector in the pulse : . *,* .. :* .. * .. *. **;*'., .*:.*;;, '. *,, ; : . ' ' position. Reactivity Worth of an Experiment: The reactivity worth of an experiment Is the yalue of the reactivity change that results from the experiment being its .. ) * * *-'* ','i ,,

• I '* Reactor Operating:

The reactor is operating whenever it is not secured or shut down. Reactor Operator: An individual licensed to manipulate "the co'htrols ofa reactor. Reactor Safety Systems: Reactor safety systems_ are those systems, including their associated ipput channels, which are designed to initiate, automatically or manually, a reactor scram.for the primary . purpose of protecting the reactor. Reactor Secured: The reactor is secured when: 1. Either there is insufficient moderator available in the rceacto_r to _attain cr,iticality or-th_ere is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection;

2. Or all the following conditions exist: a. All neutron-absorbing control devices are fully inserted or other safety devices are in their shutdown position, as required by technical specifications;

b. The console key switch is in the off position, and the key is removed from the key switch; c. No work is in progress involving; core fuel, in-tank core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rodsi and .. d. No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding one dollar. Reactor Shutdown:

The reactor is shut down if it is subcritical .by at.least one dollar in the reference

• core " -* condition with the reactivity worth of allinstalled experiments

_ir:iduded.' Reference core condition: The condition of the core when it is at ambient (cold,"i8-2S

• q"' : ' and the reactivity worth of is negligibleless than $0.01. Review: A qualitative examination of records, procedures or other documents.

Safety Channel: A safety channel is a measuring channel in the reactor safety system. Scram time: Scram time is the elapsed time between the initiation of a scram and the instant that the control rod reaches its fully-inserted position. 3 ,, . '_. *:* .. Senior Reactor Operator: An individual wtib is licensed to dire"Ctth'e Such an individual is also a reactor operator. Should, Shall, and May: The word "shall" is used to denote a requ\re'm'ent; the word "should" is used to. denote a recommendation; and the word "may" denotes permissi.on,* neithe*r a requirement nor a recommendation. Shutdown Margin: Shutdown margin shall mean the minimym sh,utdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems and will remain subcritical without further operator action*, starting from any* permissible operating condition' with the most reactive rod is in its most reactive position. Square-Wave Mode (S.W. Mode): The square-wave mode shall mean any operation of the reactor with the mode selector in the square-wave position. Steady-State Mode (S.S. Mode): Steady"state'mode shall*meah operation of the reactor with the mode selector in the manual or auto position.

• ,1,. ". Surveillance Intervals:

Allowable surveillance intervals shall not exceed the following:

1. Quinquennial

-interval not to exceed 70 months. 2. Biennial -interval not to exceed 30 months. 3. Annual -interval not to exceed 15 months. 4. Semi-annual -interval not to exceed 7.5 months. 5. Quarterly -interval not to exceed 4 months. 6. Monthly -interval not to exceed 6 weeks. 7. Weekly -interval not to exceed 10 days. Unscheduled Shutdown: An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system,*operator;error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation,*not including shutdowns that occur during testing or. checkout operations .... , ....... -_-;1. 4

2. Safety Limits and Limiting Safety

.. ,.i, _;r; 2.1 Safety Limit-Fuel Element Temperature Applicability. This specification applies to the reactor fuel. Objective. The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding shall result. Specifications.

1. The temperature in an aluminum-clad TRIGA fuel element shall not exceed 500 °C under any mode of operation.

2. The temperature in a stainless-steel clad TRIGA fuel element shall not exceed 1,150 °c. Basis. The important parameter for a TRIGA reactor is the fuel element temperature.

This parameter is well suited as a single specification especially since it can be measured. A loss of the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the moderator. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy. The safety limit for the aluminum-clad TRIGA fuel element is based on data which indicate that the zirconium hydride will undergo a phase change at 535 °C. This phase change can cause severe distortion in the fuel element and possible cladding failure. Maintaining the fuel temperature below this level will prevent this potential mechanism for cladding failure (SAR 4.5). The safety limit for the stainless-steel clad TRIGA fuel is based on data including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1,150 °C (SAR 4.5.4.1).

2.2 Limiting

Safety System Setting (LSSS) Applicability. This specification applies to thermal reactor power. Objective. The objective is to prevent the safety limits from being reached. Specifications.

1. The limiting safety system setting shall be a steady state thermal power of 1.1 MW. Basis. The limiting safety system setting is a total core thermal power, which, if exceeded shall cause the reactor safety system to initiate a reactor scram. This setting applies to all modes of operation.

In steady-state operation up to 1.1 MW, ample margins exist between this setting and the safety limits of 5 ' ' peak fuel temperature as specified in SAR 14.2.1, as long as the aluminulii 0 clad fue'ii's festrfcted to the F and G rings of the core assembly (SAR 4.S.4.1). Thermal and hydraulic calculations indicate that stainless-steel clad TRIGA fuel may be safely operated up to power levels of at least 1.9 MW with natural convection cooling (SAR 4.5.4.5).

• J ., *. *. , .. ,.: '*:'1*. *:;* ,*. *. 1(:\ :;'. , .. , ... ,'1'(*,.' . . . .' . ... ,., -.1* ' * ' r ' * ' ('. ' 6 ._,, ';*: ,*;.

3. Limiting Conditions of Operation

3.1 Reactor

Core Parameters

.;': I ,." 3.1.1 Steady-state Operation 3.1.1.1 Shutdown Margin Applicability.

These specifications apply to the reactor at all times that it is in operation. Objective. The objective is to assure that the reactor can be alftlmes and to th9t fuel temperature safety limit shall not be Specifications.

1. The reactor shall not be operated unless the shutdown margin provided by the control rods is at least $0.30 with the following conditions:

j', .* '*I a. Irradiation facilities and experiments in place and all movable experiments in their most reactive state; b. The most reactive control rod fully-withdrawn; and c. The reactor in the reference core condition where there is no 135 Xe poison present and the core is at ambient temperature. Calculations may be performed to determine a "no 135 Xe poison" reactivity condition. '** ;;, , * . . Basis. The value of the shutdown margin the reactor can be shut down from ariy condition even if the most reactive control rod should remain in the fully-withdrawn position. Since the reactor is seldom in a "no 135 Xe poison" condition, it is acceptable to perform calculations to determine the "no 135 Xe poison" reactivity condition. ,\, 3.1.1.2 Core Excess Reactivity Applicability. This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments, It applies*!for all modes of operation;

Objective.

The objectives that must be simultaneously met are to assure that the reactor has sufficient reactivity to meet its mission requirements, be able to be shut do'wn at any time, ahd not exceed its fuel temperature safety limit, Specifications, 1. The maximum available excess reactivity shall not exceed $7 .00 at reference core conditions. Basis. This amount of excess reactivity will provide the capability to operate the at full power with experiments in place and 135 Xe built up in the core. ". . , , 7 ')' '* rlr' .1.:' ;*, . -.

3.1.2 Pulse

Mode Operation Applicability. This specification applies to the energy generated in the reactor as a result of a pulse insertion of reactivity. Objective. The objective is to ensure that the fuel temperature shall not exceed 830 °C. Specifications. '.; ,. 1. The reactivity to be inserted for pulse operation shall be determined and limited by a mechanical stop on the rod, such that.the reactlvity insertlon shall not exceed $3.00. Basis. The fuel temperature rise during a pulse transient has been estimated conservatively to not exceed any fuel temperature limits with a $3.00 pulse insertion. 3.1.3 Core Configuration Limitations Applicability. This specification applies to mixed cores of aluminum-clad and stainless-steel clad types of fuel. :

• Objective.

The objective is to ensure that the fuel temperature safety limit shall not be exceeded due to power peaking effects in a mixed core. Specifications.

• • • 1' * .. :. 1. Aluminum-clad fuel shall only be loaded in the F and G r.ings of core. -:**.il 2., There shall be at least 110 fuel elements in the core (not including fuel-follow!;ld control . . ' ' .: I *. * ,. * ' " '\ .*, * ,l :* *, '. I * *: ! : i, , 1 ,-,, rods), 3. There shall not be a fuel element'in the central thimble.**'

:.' 4. Fuel shall not be inserted or removed from the core unless the reactor is subcritical by more than the calculated worth of the most reactive fuel assembly being moved. ' ' 5. Control rods shall not be manually removed from the.co.re the core has.,been .shown to be subcritical with all control rods in the full-out position. Basis. The limitation of power peakirg effects ensures that the safety limit shall not be exceeded in an operational core. Keeping aluminum-clad fuel in the F and G rings limits those fuel temperatures to safe values for aluminum-clad fuel (SAR 4.5.1.2). Keeping at least 110 fuel elements in the core helps reduce the power peaking in the core. 3.1.4 Fuel Parameters "; .) '. Applicability. This specification applies to.all tu.el . **ii ; . , , .. _,'. Objective. The objective is to maintain integrity of the fuel element cladding. Specifications. 8 "*i -**:.

1. The reactor shall not operate with damaged fuel elements, except for the purpose of locating , damaged fuel elements.

i .* ' 2. A fuel element shall be considered damaged and must be removed from the core if: a. The transverse bend exceeds 0.0625 inches over the length °of the cladding;

• ' .. b. Its length exceeds its original length by 0.10 inch for stainless-steel clad fuel or 0.50 inch for aluminum-clad fuel; c. A cladding defect exists as indicated by* release of fission :products;*

d. Visual inspection identifies significant bulges, pitting, or corrosion; and *.1, . ,.: e. 235 U burnup is calculated to be greater thar 50% ofinitial content. Basis. Gross failure or obvious, significant visual deterioration o(the fuel is sufficient to warrant

• declaration of the fuel as damaged. The elongation and bend .limits are the values found acceptable the USN RC (NUREG-1537).

3.2 Reactor

Control and Safety System 3.2.1 Control Rods Applicability. This specification applies to the function of the control rods. Objective. The objective is to the control rods are operable: Specifications. . .*I* 1. The reactor shall not be operated unless all control rods are operable.

2. Control rods shall not be considered operable if: .: .. a. Physical damage is apparent to the rod or rod drive* assembly and it does not respond ' normally to control rod motion signals; or '-*,,* ;* '*. b. The scram time exceeds 1 second for any shirhor regLflating rod or the scram time exceeds 2 seconds for the transient rod; or c. The maximum reactivity insertion rate of any shim or regulating rod exceeds $0.29 ' *' ' ' . ':' ,,'1 J*' : : .. ' per second. Basis. This specification ensures that the reactor shall be promptly shut down when a scram signal is initiated.

Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to ensure the safety of the reactor (SAR 13.2.2.2.1). 9 ;: ..

3.2.2 Reactor

Measuring Channels . . ... . '-'* . . *'* '. . . '. ' ' . . Applicability. This specification applies to the information which shall be available to the Reactor Operator during reactor operation. Objective. The objective is to specify the minimum riµmber of pp\\.fer r:neaswing channels that shall be. available to the operator to ensure safe operation of the reactor. l .. .-. : . .. . i . ., Specifications.

1. The reactor shall not be operated in the specifieg minimum numb!!r of power measuring channels listed in Table 3.1 is operable.

-, ' Table 3.1 Minimum Measuring Channels Measuring Channel Effective Mode --',, S.S. J>ulse s.w. --Power level (NPlOOO and NPPlOOO) 2 -2 Pulse power level (NPPlOOO) Power level (NMlOOO) 1 -1 Water temperature 1 1 1 ' *: . : < ',i '* . ' : ; ! '* -. ' .I .'* *.. Basis. The power level monitors ensure that the reactor power ls a_dequately for state, square wave and pulse modes of operation {SAR 7.2.3.1). The specifications on reactor power level indication are included in this section since the power level is directly related to the fuel temperature. The water temperature monitor ensures that water temperature will be kept within the specified limit. *'.1)(! 3.2.3 Reactor Safety System Applicability. This specification applies to the re_actor safety systElmch<!nn!!ls and interlocks. Objective. The objective is to specify the minimum number of reactor safety system channels and interlocks that shall be available to the operator .to _ensure safe.operation of the reactor. Specifications, 1.7 The reactor shall not be operatea unless the miniIT;urri bf safety channels described in Table 3.2 and interlocks described in Table 3.3 are operable . . *. ... ..... ',*,,*** 10 '. '! ' . Table 3.2 Minimum Reactor Safety Channels Safety Channel Function ., *"J\:*:*1**: j!", Effective Moqe . . S.S.' Pulse *s.w . Power level SCRAM @ 1.1. 11./lW(t) or less ,**.: .. 2. -.2 Preset timer SCRAM <15 sec af):er pulse initiation Console SCRAM button SCRAM '" ',:*-* .1 1 1 High voltage SCRAM on loss of nominal operating voltage 2 1 2 to the NPlOOO and NPPlOOO power channels Scram within 8 seconds upon lack of response Watchdog SCRAMs in DAC or CSC computer (one s'cram circuifper 2 2' *2 computer)

• • Table 3.3 Minim.um Interlocks

,, .. Effective Mode Interlock Function s.s. Pulse s.w. NMlOOO Power level Prevents control rod withdrawal 1 at <10-7%

• * --Transient Rod Cylinder Preyents application of air unless fully 1 --inserfed lkW Pulse interlock Prevents entering pulse mode above 1 1 --kyY Shim and Regulating rod drive Prevents simultaneous manual 1 1 withdrawal of two rods * -. circuits Shim and Regulating rod drive Prevents withdr'awal of aoy rod except. circuits Transient Rod Basis. The power level scrams provide 'protection to ensure that t.he

'ca*ri be shut down before ' ' ' lj , I * ** the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system if an urisafe or abnormal condition 'otturs. The high voltage ensures that the required power measuring have s_uttident high voltageias required for proper' functioning of their power level scrams. The interlock to prevent startup'ofthe reactor at courit rates less than 10'7% power ensures that the startup is not initiated unless a reliable indication of the neutron flux level in the reactor core is available. The interlock to prevent entering pulse mode above 1 to ensure that the magnitude of the pulse will not cause the fuel element temperature safety limits to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully , ' inserted is to prevent pulsing the reactor in the steady-state mode:Thecinterlock to prevent-withdrawal .* of the shim, safety or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive period. The interlock to prevent simultaneous withdrawal of two control rods is to limit reactivity insertion rate from the standard corii:rol rods. " * * * ' .* * ' 11

3.3 Reactor

Primary Tank Water ".: .... '. Applicability. This specification applies to the primary water of the reactor tank. Objective. The objective is to ensure that there is an adequate amount of high quality water in the reactor tank for fuel cooling and shielding purposes, and that the bulk temperature of the reactor tank water remains sufficiently low to guarantee ion' exchanger resin integrity. Specifications.

1. The reactor primary water shall exhibit the following,paqime,ters:

a. The bulk tank water temperature shall .not exceed 60 °C; b. The conductivity of the tank water shall be less than 5 µmhos/cm w,hen averaged over a one month period; c. The reactor shall Rot be operated if the taRk water level is more thaR 24 iRches below the top lip of the reactor taRk, a Rd aR alarm which is atidible to the reactor operator shall souRcl wheR the water level is too lowThe reactor shall not be operated, and an alarm which is audible to the reactor 'operator shall soun'd if the tank level is. more than 24 inches below the top lip of the tank; and, ., ! ' d. The reactor shall not be operated ifthe radioactivity of the pool water exceeds the limits of 10 CFR 20 Appendix B Table 3 for radioisotope's with half-lives

>24 hours.* * 'I -' NOTE: These specificatio.ns are not required to be met if the reactor fuel has been* removed from the tank. Basis. The bulk water temperature limit is necessary to ensure that the ion exchange resin does not undergo severe thermal degradation. Experience at many research reactor facilities has shown that maintaining the conductivity within the s.pecified .limit provides accepta,ble of corrosi,c;>n (NUREG-1537). The minimum water level of no more than 24 inches below the-top lip. of the reactor tank ensures sufficient cooling water both for normal _operation tank leak of 350 gpm for any aluminum clad fuel to cool to safe after a level . than 24 inches below the top lip of the tank) gives approximately 18 feet-4 inches of water above the ' ' .. l .. . . ... ,* 1*,f .. " . :*, . top grid plate of the core. 3.4 This section intentionally left blank.

• _i;* ... ! 3.5 Ventilation and Confinement

' , i :.1. ,' \_. '1 ! I '; / rr '0 \ *

• Applicability.

This specification applies to the operation.of the1faGHity:ve*ntilation and confinement system. . .'.' :*;:* .. ' * . ) : *.. *1 . '. : . . * ' -'

• Objective.

The objective is to ensure that the ventilation and confinement system shall be ope_ration to mitigate the consequences of possible releases of radioactive materials resulting from reactor operation. 12 Specifications.

• • .r* 1. The reactor shall not be operated unless a facility ventilation syste*/n'"is operating ahd the reactor bay pressure is maintained negative with respect to surrounding areas by at least 0.1" water pressure except for short periods oftime (not to exceed 2 hours) for system troubleshooting, maintenance and movement of personnel or equipment throL1gh open doors, provided the CAM is operating.

The normal mode ventilation system is considered operable if: a. The normal exhaust fan is operating; and ,. ,* b. The reactor bay is sufficiently confined to allow.a.piini.mum*differential P.ressure of 0.1" water column to be maintained by the normal exhaust fan. 2. The reactor bay ventilation system shall operate in the emergency mode, with all exhaust air passing through a HEPA filter, whenever a high level continuous air monitor (CAM) alarm is present due to airborne particulate radionuclides emitted from the reattcir or samples* in the* reactor bay. The emergency mode ventilation* system is *considered if:* a. The emergency exhaust fan is operating; and b. The reactor bay is sufficiently confined tp.allow a minimum differentiai of 0.1" water column to be maintained by the emergency exhaust fan. ' 3. Movement of irradiated fuel or fueled experiments with significant fission product outside of containers, systems, or storage areas within the reactor bay shall notlie performed unless the ventilation system is operating as described in TS 3.5, Specifications 1 and *2. If there is a failure of the ventilation system while movement of these materials is being performed, the material shall be placed in an appropriate location until the ventilation system is made operable . . ., . . 4. Core or control rod work that could cause a change ih reactivity bf more than one dollarshilll not be performed unless the ventilation system is operatin*g.as described

• in TS 3.5, Specifications 1 and 2. If there is a failure of the ventilation system while this work is being' performed, the material that could cause the change in reactivity, shall be placed in a11 appropriate location until the ventilation system is made operable .. * ., : ' '1: *. S. Movement of experiments within the core that cduld

a. change of total worth of more than one dollar shall not be performed unlei°ss'the.ve'ntilation system is'bperatin'g as described in TS 3.5, Specifications 1and2. If there is a failure*orthe 1.ieritilation system while movement of these experiments is being performed, the material shall be placed in an* appropriate location until the ventilation system is made operable. . . * * * 'i ' ' -. ' * * * --: : Basis. The worst-case maximum total effective dose equivalent is well below the 10 CFR 20 li'mit for individual members of the public. Thi.s has been shown to be for sc.en?rios.

the.ventilation system continues to operate during the MHA where the ventil.ati9n does ,not opera,te during . the MHA (SAR 13.2.1). Therefore, operation qfthe reactor sh9r,t periods while the r,eactor bay underpressure is not maintained because of testing or reactor bay open doors, does not compromise 13 ',*,,, ... 'i., . r -';. ',<<.: the control over the release of radioactive material to the unrestricted area nor should it cause occupational doses that exceed those Ii mi.ts given in ,lOSFR 29.,". 3 .6 This section intentionally blank. 1; 1 3.7 Radiation Monitoring Systems and Effluents'**'

• ** .. I 3. 7.1 Radiation Monitoring Systems Applicability.

This specification applies to the radiation monitoring systems. Objective. The objective is to specify the minimurri radiation rrioriitoring'chaririels that shall be available to the operator to assure safe operation of the reactor. Specifications.

1. The reactor shall not be operated unless numper monitoring

.. channels listed in Table 3.4 is operating. Each for the Environmental, Dosimeters shall have a readout in the control room and be capable of sounding an audible alarm. Table 3.4: Minimum Radiation Monitoring Channels* Radiation Monitcfring Channel' Number' Continuous Air Monitor sampling reactor bay air* *: .. 1 Radiation Area Monitor in reactor bay Environmental Dos.i'meter outside reactor facility 3 41 Ar Monitor sampling stack exhaust '. ;, 1 0 The Continuous Ai'r Monitor or the may be out-of-service fqr up to 2 hours.for caliqration, troubleshooting, or repair. During this out-of-service* time, no experiments or maintenance activities sliall*be conducted which could directly result in alar'r'ri conditions (e.'g.,'airborne releases or high radiation levels), and the ventilation be ' operating. A portable, gamma-sensitive ion charnber, with display . .' , . *. . *. .-. ! . .' ' '

• l
• visible from the c,ontrol room, may be utilized as .a tenippra.ry., substitute for the Radiation Monitor (put. not for the Continuous Air Monitor) for a period Up*to 60 days. " .. Calculations may be performed to determine 41 Ar releases as a* function of reactor operating history as a temporary substitute for the required 41 Ar monitor for a period up to 60 ,days ... ) ,, * \. * . I * . *-:p '.' of 1 *'I *:*J

  'ljl ,-: . , ) i . . \ : ., 1 I * * *.:; :.* "'.:' r1*L:; ;; ** Basis. The radiation monitors provide ihformation to operatirl{{p'e'rs'Bnriel regardi'ng routine of radioactivity and any impending or exi'stin'g'danglfr frb*m* radiation.The alarm setpoints are chosen to be at levels higher than those normally encountered during routine reactor'operations:

Their operation Will 14 provide sufficient time to evacuate the facility or take the spr,ead .of ... , radioactivity to the surroundings (SAR 11.1.6). 3.7.2 Effluents Applicability. This specification to the rel.ease rate of 41 Ar. ;,;**-* J (* .. Objective. The objective is to ensure that the concentration of'the 41 Ariil 'the areas ;hall be below the applicable effluent concentration in 10 CFR20. I l*' '*' Specifications. l, The annual average concentration of 41 Ar discharged into the unrestricted area shall not. exceed 4.8 x 10*5 µCi/ml at the point of discharge. Basis. If 41 Ar is continuously discharged at 4.8 x 10*5 µCi/ml, measuremen.J:s. and calculations show that 41 Ar released to the publicly accessible areas under the worst-case weather conditions would result in an** . annual TEDE of 0.5 mrem. This is only 5% of the applicable limit of 10 mrem. The calculation was performed with the Environmental Protection Agency's Comply code (SAR 11.1.1.1.4).

3.8 Limitations

on Experiments

3.8.1 Reactivity

Limits Applicability. This specification applies to experiment's Installed in the reactor.and its'irradiation f .l,_, . . *l ' ... : facilities. Objective. The objective is to prevent damage.to.the reactor or.excessive release of'radioactive materials in the event of an experiment failure;* Specifications.

1. The reactor shall not be operated unless the foUowing conditio.ns governing.

experiments, exist: a. The absolute reactivity worth of any be less than $1.00; and b. The absolute reactivity worth of any single secured experiment shall be less than $3.00; and .' ,. c. The sum of the absolute reactivity worth for ali experiments be less '$5.00.' . . *' ' .-Basis, The worst event which could possibly arise is the sudden r.emoval of a..movable experiment immediately prior to, or following, a pulse transient of the maximum licensed reactivity insertion. Limiting the worth of a movable experiment to less than $1.00 will ensure that the additional increase of transient power and temperature is slow enough for the high power scram to be effective and, since this* '.-. 15 : .-. :' ': : * .. transient is not a super-prompt pulse, we would not'violatethe'-i kW Pulse Interlock whkh prevents entering pulse mode above 1 kW (SAR 14.3.2.3). The worst event that is considered in conjunction with a single secured experiment is the sudden removal of the experiment while the reactor is operating in a critical condition at a low power level. Thi's

• is equivalent to pulse-mode operation qfthe reactor. Hi!nc;:e,,the,reactivity limitation for a si,ngle secured experiment at $3.00 is the same as of a maximum a scram would be **'. initiated much more quickly for the experiment removal accident (SAR 13.2.2.2.1and14.3.1.2).

3.8.2 Materials

Applicability. This specification applies to.experiments installed.in the and its irradiation

• facilities.

Objective. The objective is to prevent damage to the reactor or excessive release ofradioactive materials in the event of an experiment failure. Specifications.

_

• I ' ' 1. The reactor shall not be operated unless the following conditions governing experiments exist: a. Explosive materials, such as gunpow9E!r, TNT, or nitrogly_cerin, in greater .*. _ than 25 milligrams TNT-equivalent shall not be irradiated irradiation

,.,, -* facilities. Explosive materials in quantities less than or equal to 25 milligrams TNT-equivalent may be irradiated provided the pressure produced upon detonatiorrofith'e explosive has been calculated and/or experimentally demonstrated to be Jess than' halhi* _ ... ** *

• the design pressure of the container;

b. Each fueled experiment shall be controlled such that the total inventory of 131 1-135 1 in the experiment is no greater than 1.5 the total inventory of 90 Sr in the experiment is no greater than 5 millicuries; and c. Experiments containing

'inateriafs shall be doubly encapsulated."The . of an encapsulation of material that could damage the reactor shall result in removal of the sample and physical.i_nsP,ection of poter;itia}ly pamaged components_. Basis. This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials (S.A.R 13.2.6.2). The 1.5-curie limitation on 131 1-135 1, and the 5 millicurie limit on 90 Sr, ensure that in the event of a involving total release of the iodine, the dose in the reactor bay and in the *unrestricted area will b*e considerably less than that allowedbylOCFR20(SAR13.2.6). ,,_,,,. * '*'*'"*:":.*: , .. ,, .,. 3.8.3 Failures and Malfunctions . :*** Applicability. This specification applies to experiments installed in the reactor and its irradiation facilities. 16 Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure. Specifications. . , . 1. Where the possibility exists that the failure of an experiment (except fueled experiments) under normal operating conditions of the experime*nt or .conditions in the reactor, or possible accident conditions in the experiment could release radioactive gases or aerosols to the reactor bay or the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in ttie reactor bay or the unrestricted area will not result in exceeding the a'pplicable dose limits in: 10 CFR 20, assu'ming that: a. 100% of the gases or escape from the exper.iment;

b. If the effluent from an irradiation facility exhausts through a holdup tank which closes automatically on high radiati.on I eve!, a.t leasUO% of tbe activity or aerosols.

produced will escape; c. If the effluent from an irradiation facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 at least 10% of these aerosols can escape; and d. For materials whose boiling point is above 130 °F and where vapors formed by boilirig this material can escape thrciu.gh an undisturbed'.column of water above the 10% of these vapors . ' . Basis. This specification is intended to meet the purpose of 10 CFR 20 by reducing the likelihood that released airborne radioactivity to the reactor bay, or unrestricted area surrow:iding the GSTR will result in exceeding the total dose limits to an indiyidual as specified in 10 CF.R 20. . * * * 'f' 3.9 This section intentionally left blank. ,,* '; \': !',:* 17

4. Surveillance Requirements

' \ ; 4.0 General Applicability.

This specification applies to surveillance requirements of systems related to reactor safety. Objective. The objective is to verify the operability of to reactor safety. . ' ' * -* ; -' ;*J' ' ' ': ' : ' ; -' Specifications.

; 1. Surveillance requirements may be deferred during, reactor shutdown (ex.cegt TS 4.3 Specifications 1 and 3, and TS 4.7 1, 3, an,q 4). However, they soall be completed prior to reactor startup unless reactor operation is required for performance of the surveillance.

Such surveillance shall be performed as soon as practical after reactor startup. Scheduled surveillance which cannot be performed withtlie reactor operating may be*d'eferred until a planned reactor shutdown. .. * ! 2. Any additions or modifications' to the ventilation

• system, the core and its associated support structure, the pool or its penetrations, the primary coolant system, the rod drive mechanism or the reactor safety system shall be made and tested to assure that the systems will meet their . . . . . . . functional requirements in accordance with marufac!urer specifications or specificatio.ns reviewed by the ROC. A system shall not be considered operable until after it is successfully.*.:*

tested. , . ;: ' * * ' . . I ; " 3. The reactor control and safety systems, pool water level alarm, and radiation monit'oring

• systems shall be tested to be operable after the completion of respective items. Basis. These specifications relate to change's in reactor could affect t.he safety of th< reactor. These changes will be formally addressed by following the requirements of 10 CFR 50.59. *As long as changes or replacements to these systems meet or exceed the original design then it can be assumed that they meet the presently accepted operating criteria.

Additional requirements may be needed, based on the evaluation through the 10 CFR S0.59 process. This specification is not intended to circumvent or replace the regulations in 10 CFR 50.59. 4.1 Reactor Core Parameters Applicability. This specification applies to the surveillance requirements for reactor core parameters. Objective. The objective is to verify that the reactor does not exceed the authorized limits for power, shutdown margin, core excess reactivity, specifications for fuel element condition and verification of the total reactivity worth of each control rod. Specifications. 18 . ,.*1 i !

1. A channel calibration shall be made of the power level monitoring channels by the ... calorimetric method at least annually.

2. The total reactivity worth of each control rod shall be measured annually or following a change in core or control rod configuration that is expt?cted to the

_reactivity worth of that control rod by more than $0.30 (not including transient fission p.roduct poison

3. The maximum reactivity insertion rate of a stantlar-d-control rod shall be measured annually or following a change in core or control rod configuration that is expected to change the total reactivity worth of that control rod by .more than $0.30 (not.including.transient fission product poison effects).

4. The core shutdown margin shall be determined at least annually and following a change in core or control rod configuration that is expected to change the shutdown margin by more than $0.30 (not including transient fission' pro*duct poison effects).

.,,, '. *, '* * . ,*, : . ; } ,., . i . . " . i . 5. The core excess reactivity shall be determined following a change in core or

• control rod configuration that is expected to change _the excess reactillity by *more than $0.30 . (not including transient fission product poison effects).

6. The transient rod and drive mechanism shall be.tested and inspected at least annually.

r; , 7. Verification of core configuration to include aluminum-clad fuel only in the F and G. rings of the core and to have a 1 of llO elements in the s 0 hall be by visual . . . ' ' . . means prior to each day of operation.

-., ,(, ' . .

* . ' . . ' 8. All fuel elements shall be inspected for damage or deterioration and measured for length and transverse bend at least at qµi.nquennial intervals or if 500 pulses have been .performed since . the last fuel inspection.

9. For each month during which pulsing is performed, the relationship between peak fuel temperature and inserted reactivity.shall be determined.

NOTE: These checks are not required if reactor fuel has been removed from the tank. Basis. Experience has shown that the identified frequencies will ensure perform?nce and operability for each of these systems or components. Movement of the core components could change the reactivity of the core and thus affect both the core excess reactivity and the shutdown margin, as.well as.affecting the worth of the individual control rods. Evaluation of these parameters is therefore.required after any such movement. Without any such movement, the changes of these over' an period oftime and operation of the reactor have been shown to be small, so that an annual measurement is sufficient to ensure compliance with the specifications. Experience at TRIGA reactors indicates that examination of a five-year cycle is adequate to detect problems; A five-year cycle reduces* the handling of the fuel elements and thus reduces the risk of accident or damage due to handling. 19 :. ' '. .

• j :r,*. . l*'

4.2 Reactor

Control and Safety Systems** 1

• Applicability.

This specification applies to the surveillance requirements of reactor control and safety systems. Objective. The objective is to verify performance and operability of those systems and components ' . : '* . which are directly related to reactor safety. Specifications. ' ., . 1. The control rods shall be visually inspected for damage.or deterioration' at least biennially.

2. The scram time shall be measured at least annually or after any work (not including routine limit switch adjustments) is performed on: a control rod,drive.

3. A channel test of each of the reactor'safety system channels in.1cible 3.2 for the mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day. The same channel tests shall 'be performed after modifications or. repairs to the scram channels to ensure operability of the respective channels.

- 4. A channel test of items in Table 3.2 and 3.3 shall be performed at least semi-annually, except for those two items required solely for pulse mode operation, which shall be channel tested during each startup for pulse mode operation. The two items required solely for pulse mode ' * " I ;* ' 1 , ' " I ' * : ") '.: ', , '.* .. / operation are the Preset timer scram in Table 3.2 and the .control rod interlock in Table 3.3 that' prevents withdrawal of any rod the Rod. . "' ' :*' * * ',,;*._ (H NOTE: These specifications are not required if the reactor fuel has been removed from the tank. -* * ' * * * * =' **:* .' n; ;;.c;.

• Basis. Inspection of the control rods allows early detection of signs*of deterioration indicated by sigriSiof

!r. :. changes of corrosion patterns or of swelling, bending, or elongation. The channel checks performed daily before operation and after any modifications or repairs provide timely assurance that the systems will operate properly during o*peration of the reactor. Experience has shown that the identified frequencies will ensure performance and operability for each : of these systems or components.

l. ' ' 4.3 Reactor Primary Tank Water Applicability.

This specification applies to the surveillance requiren'i*nts for the reactor tank water. Objective. The objective is to ensure thatthe reactor tank wate(level arid the bulk water temperature monitoring systems are operating and 'to verifyappropriate'a1ar'm'set1:ings'.

• ' .... v Specifications.

  1:1 ;_,,; ,;:,,: '.*'* ! 1. A channel test of the reactortank water level alarm:setpoint Shall be performed at least semi-annually.

• 1" ,; ' 20

2. A channel check of the reactor tank bulk water temperature alarm setpoint shall be performed quarterly.

A channel calibration of the reactor tank bulk water shall be performed at least annually.

3. The reactor tank water conductivity shall 'be measured ino"nthly. -Multiple mea;,u;ein.ents . taken in one month shall be averaged to determine the monthly value. 4. The pool water radioactivity shall least NOTE: These specifications are not required if the reactor fuel has been removed from the tank. Basis. Experience has shown that the frequencies of checks on systems which monitor reactor primary water can adequately keep the tank wate'r at the proper 1eJe1

'ma'lntain water qu.ality at such a ievel. to minimize corrosion and maintain safety. Experience at the GSTR shows that the surveillance specification on the conductivity is adequate to detect the onset of degradation. of the quality of the pool water in a timely fashion. Experience also indicates thafthe su*rveillahce specification 'on* pool water level and pool water temperature are adequate to detect losses of pool water in a timely manner and to enable operators to take appropriate action ::Viien the coolant the .. specified limit. The quarterly and annual surveillances of the temperature monitor are also adequate to assure operability of the temperature channel. The pool water level alarm system is a reliable unit and therefore the specification of a semi-annual test is sufficient to assure operability of the pool wai:er level alarm. 4.4 This section intentionally le'ft blank. 4.5 Ventilation and Confinement System Applicability. This specification applies to the reactor bay ventilation and confinement system. Objective. The objective is to operation of the ventilation an*d confine'ment system in* 1

• controlling releases of radioactive material to the unrestricted area. Specifications.

,, ', .. *)_ 1. A channel check of the reactor bay ventilation shall be performed prior to each day's operation or prior to each operation extending more than one day. 2. A channel test of the reactor bay.,veritilation system's ability to'*automatically switch to the* emergency mode upon actuation of the CAM high alarm and td proliid 1 e a reactor bay minimum differential pressure of 0.1 11 water column shall be performed quarterly. . . ' Basis. Experience has demonstrated that checks of the syste'm on the prescribed'fr'equenCies are sufficient to ensure proper operation of the system and its df radioactive material. 21 .... t ** 4.6 This section intentionally left blank. * * "*°'.'. .* * ,i, 0:1.1.*** : *. ' . ; :.. .. *. . '* . 4. 7 Radiation Monitoring System Applicability. This specification applies to the surveillance requ_irements for the area radiation monitoring equipment and the air monitoring

• * . .. .. !:*.!: Objective.

The objective is to ensure that the radiation monitoring equipment is operating properly and to verify the appropriate alarm settings. ';' ;*, > ' !' *.:: *. *. Specifications. ,._., :(: * *. ' .* .** ' ., ,,. ! '* *. , . ) I 41. ..* .. . 3 ' 1. A channel check of the radiation are<! air monitor, and ,monitor shall be performed monthly. *'*,, ;*. 2. A channel test of the air ' , 1 , , , * ' ' i} ' ' 'I ' : ( , .

• 1 , ' 3. A channel calibration of the radiation area monitor arid' continuous air monitor and 41 Ar monitor shall be performed

.,, . *** .,.:*. 4. The environmental dosimeters

• s'hall be changed and at least annually: . , , I ' Basis. Experience has shown that an annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span. The frequency of changing and evaluating environmental dosimeters are also adequate to provide the required record b9se?;qn,., past experience.

. ,, . i *. . .. .*. f{: . . i 4.8 Experimental Limits Applicability.

This specification appli17s to surveillC!nce for experiments . .,,,: reactor and its irradiation facilities.

• Objective.

The objective is to prevent the conduct of experiments which may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure. . Specifications.

1. The reactivity worth of an shall be or n;i eas\,lred, as appropriate 1

.. routine reactor operation wit,h that exp,eriment to limits of TS 3.8.1 are n_o.t exceeded. 1i. 1 ** ,,* *** 2. An experiment shall not be in the or its a analysis has been performed and for ... !:; ,3.8 1 ,2 and TS 3.8.3 by,the, . Reactor Supervisor or ROC in full accord with TS 6.2.3, and the procedures which are established for this purpose. 22 Basis. Experience has shown that experiments which are reviewed by the staff of the GSTR and the ROC can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.

4. 9 This section intentionally left blank. ',] ,,; .;_ . ;, , . 23 ,*,*

5. Design Features 5.1 Site and Facility Description Applicability.

This specification applies to the U.S. Geological Survey TRIGA Reactor site location and specific facility design features. Objective. The objective is to specify the location of specific facility design features. Specifications.

1. The licensed area shall be the following locations on the Denver Federal Center: a. Building 15: Rooms 149 through 152, Rooms 154, 157, 158, BlO, BlOB, and Bll; b. Area inside the wrought iron fence and south cooling tower wall that is near the SW corner of Building 15; c. Building 10: Room 2. 2. The reactor bay volume shall be a nominal 12000 cubic feet and shall be designed to restrict leakage. 3. The reactor facility shall be equipped with a ventilation system designed to exhaust air and other gases from the reactor bay and release them from a vertical level at least 21 feet above ground level. 4. Emergency controls for the ventilation system shall be located in the reactor control room. Basis. The reactor building and site description are strictly defined (SAR Chapter 2). The facility is designed such that the ventilation system will normally maintain a negative pressure in the reactor bay with respect to the outside atmosphere so that there will be no uncontrolled leakage to the unrestricted environment.

Controls for normal and emergency operation of the ventilation system are located in the reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with minimum exposure to operating personnel (SAR 9.1 and 13.2.1). 5.2 Reactor Coolant System Applicability. This specification applies to the tank containing the reactor and to the cooling of the core by the tank water. Objective. The objective is to ensure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding. Specifications.

1. The reactor core shall be cooled by natural convective water flow. 24 Basis. 2. The tank water inlet and outlet pipes to, \_he he.at

apd (_o}he shaJlb_e 1 ,. equipped with siphon breaks 14 feet above the top of the core or higher. NOTE

These specifications are nqt required be met if .the core has been defueled.

1. This specification is based on thermal and hydraulic calculations which show that the TRIGA core can operate in a safe manner at power levels up to 1.9 MW with natural convection flow of the coolant water (SAR 4.5.4.5).

2. In the event of accidental siphoning of tank water through'inlet and outlet pipes o'fth'e heat exchanger or demineralizer system, the tank water. level will drop to a level no less than 14 feet from the top of the core (SAR 5.2). 5.3 Reactor Core and Fuel :. .. t*.**:. 5.3.1 Reactor Core (,I ' Applicability.

This specification applies to the configuration of and experiments. Objective. The objective is to ensure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance th at' excessive power densfties shall not be produced. Specifications. . .,-. 1. The core shall be an arrar:iger:nent,of.TRIGAuranium-zirconium hydride fuel-moderator elements positioned in the reactor grid plate:* 2. The TRIGA core assembly shall consist of stainless-steel clad* fuel elements (8.5 tq 12.0 *wt% uranium), aluminum-clad fuel a thereof. * .. * * .* .... *

• _!*, '* *>-' .. Basis. 3. The fuel shall be arranged iri a dose-packed configuration' except for element positions occupied by in-core irradiation facilities; graphite.dummies, aluminu'rii dummies,*;

stainless steel dummies, control rods; and startup sources. The core may also contain two " separated experiment positions in the D through E rings, each occupying a* maxim.um of three fuel element positions.

• • {. 4. G-ring grid positions may be empty {water filled). :-. ..i 5. The reflector, excluding experiments and irradiation facilities, shall be graphite, water, or a combination of graphite and water: A. reflector is not requtred'ifthe core has been *de'fueled'.

.. 1. Standard TRIGA cores have been in use for years and their characteristics are well documented. Analytic studies performed at GSTR for a variety of mixed fuel arrangements 25 f':* '. :*t* ;!. indicate that such cores with 'mii<e'd loadings would safely satisfy all 'operational requirements (SAR 4.2). ' "'. .. 2. The core will be assembled in the reaetor* grid plate which is lo'cated in a tank of light water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of irradiation facility radiation requirements (SAR 4.2). 5.3.2 Control Rods . , ' i Applicability. This specification applies to the control rods used in the reactor core. Objective. The objective is to ensure the control ro9s of such ,a as. to pern:iit their use ... with a high degree of reliability with respect 1 t9 th.eir physical and nuclear. characteristics. Specifications.

1. The shim and regulating control rods shall have scram capability and contain borated graphite, B 4 C powder or boron, with its compounds in solid form as a poison, in aluminum or stainless steel cladding.

These rods may incorporate fueled followers. , 2. The transient control rod shall have scram capapility and cqntajn graphite, B 4 C. powder or boron, with its compounds in a solid form as a, poison an aluminum or stainless steel cladding. The transient rod drive mechanism shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an aluminum-or air-follower. Basis. The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B 4 C powder or boron with its compounds in a solid form. These milterials*must be contained in"'

• a suitable clad material such as aluminum or stainless steel to ensure mechanical stability during ** .. :*,r*ci1 ,,, * ** movement and to isolate the poison from the tank water environment.

Control rods {that are fuel-followed) provide additional reactivity to the core and of the controi' rod.'The of fueled-followers has the additional advantage of reducingflu'x the water-filled regions vacated by the withdrawal of the controlrods. Scrarri provided f_qr rapid i,nsertion of the: control rods which is the primary safety feP,ture of the transjent control rod is for rapid withdrawal from the reactor core which rrsults in a P.ulse, The nuclear behavior. of the air-or aluminum-follower, which may be into trans\ept rod, i_s similar to a.void. A more detailed description of the control rods and their properties can be found in SAR 4.2.2. 5.3.3 Reactor Fuel Applicability. This specification applies to the fuel elements used in the reactor core. Objective. The objective is to ensure that the fuel elemE!nts are. of, a design and fabricated in.such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics. Specifications. 26 1.1:"' Basis. 1. Aluminum-clad TRIGA fuel. The individual unirradiated aluminumcclad fuel elements shall . ' . * * * ' ' * * * ; * , ",;'._. ) ' " , . * . " . * * ' '! . have the following characteristics: . . . '" *c:;i ,r, .. * .. : , . ... .. a. Uranium content: nominally 8.0 wt% with a mu enrichment. of less than 20%; ..

• t* b. Hydrogen-to-zirconium atom* ratio riominaliy 1 to 1;' and : * ..:*-*.; c. Cladding is aluminum of a nominal 0.030 inch thickness.

2. Stainless-steel clad TRIGA fuel. The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

a. Uranium content: nominal range of 8.5 to 12.0 wt% with a mu enrichment of less than 20%; b. Hydrogen-to zirconium atom ratio nominally between 1.6 to 1and1.7 to 1; and c. Cladding is 304 stainless steel of a nominal 0.020 inch thickness.

1. A nominal uranium content of 8 wt% in an aluminum-clad TRIGA element is less than the traditional stainless-steel clad element design value of 8.5 wt%. Such a decrease gives a lower power density. The nominal hydrogen-to-zirconium ratio of 1to1 could result in a phase change of the ZrH if fuel temperature is allowed to exceed 535 *c. Although this would not necessarily cause a rupture of the fuel cladding, it would cause distortion and stressing of the cladding.

2. A maximum nominal uranium content of 12 wt% in a standard TRIGA element Is about 50% greater than the lower-loaded nominal value of 8.5 wt%. Such an increase in loading would result in an increase in power density of less than 50%. An increase in local power density of 50% reduces the safety margin by, at most, 10%. The maximum hydrogen-to-zirconium ratio of 1.7 to 1 could result in a maximum stress under accident conditions to the fuel element cladding of about a factor of 1.5 greater than the value resulting from a hydrogen-to-zirconium ratio of 1.6. However, this increase in the cladding stress during an accident would not exceed the rupture strength of the cladding.

5.4 Fuel Storage Applicability. This specification applies to the storage of reactor fuel at times when it is not in the reactor core. Objective. The objective is to ensure that fuel which is being stored shall not become critical and shall not reach an unsafe temperature. Specifications. 27

1. All fuel elements and fueled devices shali'be stbred arraywhefe the effective is less than 0.9 for all conditions of moderation and reflection.

2. Irradiated fuel elements and stored in an array which will sufficient natural convection cooling by water or air the temperature ofth,e fuel element or fueled device will not exceed design values. t ;, 3. If stored in water, the water quality shall be maintained according to -TS 3.3, Specification 1.b. : .'* .. Basis. The limits imposed are conservative and ensure safe storage (NUREG-1537).

'.!.'-l " * * ' ..... ,., ;i .* .. (. 28 . .. ,:

6. Administrative Controls .,; : ' : . _.*;* 6.1 Organization Individuals at the various management levels, in addition to being responsible for the policies and operation of the reactor facility, shall be responsible for safeguarding the public.and facility personnel from undue radiation exposures and for adhering to all requirements ofthe operating license, technical

• specifications, and federal regulations.

The minimum q'ualification for all members of the 'reactor operating staff shall be in accordance with ANSI/ANS 15.4, "Selection and Training of Personnel for Research Reactors." ' ; 6.1.1 Structure The reactor administration shall be related to the USGS structure as shown in Figure 1. 6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be with the cha_in_of command* established in Figure 1. Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shaJJ *be-responsiqle.for safeguarding the public and facility personnel from undue radiation exposures and for a.JI requirements of the operating license, the established charter, and the technical specifications: The following specific organizational levels and responsibilities exist: . 1. Reactor Administrator (Level 1): The Reactor Administrator is responsible to \he US.GS. Director and is responsible for guidance, oversight, and management support of reactor operations;

2. Reactor Supervisor (Level 2): The Reactor Supervisor reports to the Reactor Administrator and is responsible for directing the activities of the Reactor Operators

ind Senior Reactor Operators and for the day-to-day operation and maintenance of the reactor; 3. Senior Reactor Operator-in-charge (Level 3)

The Senior Reactor Operator-in-charge reports to the Reactor Supervisor.

This person is primarily involved in the oversight.and direct.

• manipulation of reactor controls, oversight and direct operation and maintenance of reactor related equipment, and oversight of recovery from unplanned shutdown*s; and' 4. Reactor Operator (Level 4): Other Senior Reactor'Opera.tors a'n'a Reactor Operators report to Senior Reactor Operator-in-charge and the Reactor Supervisor and are primarily involved in the direct manipulation of reactor controls, monitoring of and direct _oi:ieration and maintenance of reactor-related equipment.

29 ,_1. ' Director U.S. Geological Survey Reactor Administrator (level 1) Review/Au!lit . . . . Reactor Operations Committee I I I :.:.* Reactor Health Physicist 1-----J Reactor Supervisor (!eve) 2) I I I ________ J "{; I I I I I I '--------------- Senior Reactor Operator-in-charge (level 3) Reactor Staff(level

4) Senior Reactor Operators Reactor Operators

• • • • • .

• • ... *.-... i.* *

    :,*:t t .. :: . *::: Line of Responsibility Line of Communication

---

*

i<" Figure 1: Administrative Structure

6.1.3 Staffing

... ,'. **r:i 1. The minimum staffing when the reactor is not secured shall *be: .. ; a. A Licensed Operator in the control room; /,"' ... b. A second -person presen,t with.in the Denver federal who is able to carry out prescribed instructions; . \ *, c. If neither of these two indivfduals* is a s'enior Reador Operator; a Senior Reactor Operator shall be readily available on call. Readily available on call means an individual who: i. Has been specifically designated and the designation is known to the operator on duty; ii. Can be contacted by phone, within 5 minutes, by the operator on duty; and 30 iii. Is capable of getting to the reactor facility within a reasonable time under normal conditions {e.g., 30 minutes or within a*

• * * * . d. It is not neGessary to have a SRO on call if the Reactor Operator in the rnntrol room is a SRO. If the Reactor Operator in room is a SRO; a sernnd person sha.11 be . *' available at the facility or on call; :tlnd ... ' * :** *'.. ed. A list of management personnel, radiation personnel, and reactor staff along with their contact information shall b'e availabie to the o'n duty: 2. Events requiring the direction of a Senior Reactor Operator:

a. Initial approach to critical after each completedshytdo)Nn

-. b. Initial approach to power after each completed shutdown checklist;

c. All fuel or control rod relocations within the reactor core region; d. Relocation of any in-core components

{other than normal control rod movements) or experiment with a reactivity worth greater than one dollar; or e. Recovery from an unscheduled shutdown.or significant {>50%),po)Ner reduction.

6.1.4 Selection

and Training of Personnel The selection, training and pf ope.rations persol")nel shall follow the guid;mce, o.fANSl/A!'JS 15.4, "-Selection and Training of Personnel for Research Reactors." 6.2 Review and Audit The ROC shall meet at least semi-annually for the purpose of providing their primary responsibility of review and audit of the safety aspects of reactor facility operations.

6.2.1 Composition

and Qualifications The ROC shall be composed of at least four voting members, induding the Chairman. All members'of the Committee shall be knowledgeable in subject matter related to reactor operations.- To expedite Committee business, a Committee Chairman may-shall be appointed. . The Chairman of the ROG shall be listed by name on the Committee roster. The Committee shall be appointed by the USGS Director. -No definite term of service shall be specified; but should a vacancy occur in the Committee, the Director shall appoint a replacement. -The remaining members of the Committee shall be available to assist the Director in the selection of new members.* ,, The Reactor Supervisor shall be an ex-officio member of the Committee;-and th.e*Reactor Supervisor" shall be the only non-voting member of the Committee. The ROC shall report to the Reactor Administrator. 31 ., .. ' !./,:*: ' .. *:** . \ .. \,'.' *,.-, . : :

6.2.2 Charter

and Rules . . . . .. , . , . '* _1* , * : , , ; _. .

• The ROC consists of USGS members and non-USGS a,nd.the Coml'1;1ittee meet at least semi-annually.

The review and audit functions shall be conducted in accordance with an established charter for the ' . ' Committee as written in the USGS Manual. Dissemination and review of Committee minutes shall be done within 60 days of each respective Committee meeting. A quorum for review, audit, and approval purposes shall consis_t nqt less.than one-half of the voting membership where the operating staff does not constitut'e a majority. The Chairperson or.an alternate must be present at all meetings in which the official business of the committee is being conducted. Approvals by the committee shall require an affirmative vote by a majority of the non-USGS members present and an affirmative vote by a majority of the USGS m'ernhers . 6.2.3 Review Function The following items shall be reviewed: . 1. Determinations that proposed*changes.'in the fatility, arid procedures,'and the condJct.of tests; or experiments are allowed without prior authorization by the res130nsible a1o1tl:!orityNRC, as detailed in 10 CFR 50.59; 2. All new procedures and major revisions thereto having safety significance, prbposei:l 2hanges':;

in reactor facility equipment, or systems having safety significance;

3. All new experiments or classes of experiments that could have reactivity or safety significance;

• ,*_;, 4. Proposed changes in technical specifications, license, or charter; 5. Violations of technical specifica.tions,.

license, or cha.rter. of internal procedures.or instructions having safety significance;

6. Operating abnormalities having safety significance;

7. Reportable occurrences listed in TS 6.7.2; and 8. Audit reports. *i A written report or minutes of the findings and recommendatiOns ofthe review shall be submitted*to the Reactor Administrator and the ROC within 3 months aften.the review has been completed:

.... 6.2.4 Audit Function 32 The audit function shall include selective (but comprehensive) examination of other documents, and the reactor facility. Discussions with cognizant personnel and observation of operations should be used also as appropriate. In no case shall the individual immediately responsible for the area perform an audit in that area. The items shall be audited: _:!,I .' 1. Facility operations for conformance to the technical specifications and applicable license eF *'"I * -, . ,, * *' *.**' "f * , * . ffiartef-conditions: at least once per calen'dar year' (interval between au'clits hot to exceed 15 . months); ,,*, 2. The retraining and requalification program for the operating staff: at fe*astonce every other * ' calendar year (interval between audits not to exceed 30 months); , . ' . . 3. The results of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or of. tha_t a:ffect reactpr safety; at once per calendar year (interval between audits n,ot to exc'i!ec:)

4. The reactor facility emergency plan, implementing procedures;'

an'd-security plan: at l'east: once every other calendar year (interval between audits not to exceed 30 months). Deficiencies uncovered that affect reactor safety shall immediately be reported to the Reactor Administrator. A written report of the find(ngs of the audit shall be submitted to the Reactor Administrator and the ROC within 3 months after the audit has beeri' completed.

6.3 Radiation

Safety The Reactor Supervisor, in coordinati_on

with, shall be responsible for_ implementation of the radiation safety program. The requirements ofthe radiation safety program are established in 10 CFR 20. The program should use the guidelines of the ANSI/ANS 15.11-2009, "Radiation Protection at Research Reactor Facilities." -_,-,. ::. 6.4 Procedures Written operating procedures shall be reviewed, and approved prior to initiating any of the ;*

• activities listed in this section. The procedures shall be reviewed by the HOC and.approved by the . Reactor Supervisor, and such reviews and approvals shall be documented ih a timely manner. Substantive changes to the procedures shall be made effective only after documented review by the ROC and approval by the Reactor Supervisor.

Minor modification to the original procedures that do not change their original intent may be made by the Reactor Supervisor. Temporary deviations from the procedures may be made by the responsible SRO or Reactor Supervisor i_n order to. deal with.special or , ; . unusual circumstances or conditions. Such deviations shall be documented and reported within 24 hours or the next working day to the Reactor Supervisor. Procedures shall be in effect and in use for the following items: "" ". * : "* 1. Surveillance checks, calibrations, and inspections that are required by Technical Specifications or those that may have an effect on reactor safety;-;_1 **: 2. Startup, operation and shutdown of the reactor; 33 -,*_.*

3. Implementation of emergency and security plans;'

• 4. Core changes and fuel movement;, .. 5. Performing maintenance on major components that could affect reactor safety; ' 6. Administrative controls for lllairitena,r:ice, c9ul.d reactor safety or core reactivity;

7. Radiation protection, including A LARA requ.irements; 8. Use, receipt and transfer of licensed radioactive materiai, 'if appropriate.

• 6.5 Experiment Review and Approval . 1. All experiments proposed for itie either 1'or 'class Ii experiments and shall be reviewed in accordance with the 10 CFR SO.S9

're'c{ufrements: The revieW and classification of the proposed exp,erir:nents .the. of Reactor Supervise'.*

2. Class I experiments include all that have been run previously or that are' minor modifications to a previous experiment.

These are experiments which \nvolve small changes in reactivity, no external shielding changes, and/or limited amounts of ra!;Jioisotope The Reactor Supervisor has the authority to approve the following:, .

• a. Experiments for which there exists adequate precedence for assurance of safety; b. Experiments which represent less than that amount of worth

'fQ'r*

• prompt criticality; or )°' '. ;,'t r ** !!H;':f('itr

c. Experiments in which any significant reactivity worth is stable and mechanically fixed','i

'..*:: that is, securely fastened or bolted to the reactor structure.

3. Class II experiments include all new exJJeriments and major modifications of previous experiments.

These experiments must he reviewed and approved by the ROC before being run., These experiments may involve larger changes in reactivity, external shielding changes, and/or., larger amounts of radioisotope production.

• . 1 i*' 6.6 Required Actions 6.6.1 Actions to Be Taken in Case of Safety Limit'\Tiofatiori In the event a safety limit is exceeded: . *I.:*:*_..

1. The reactor shall be shutdown and reactor operation shall not be resumed until authorized by the NRC; 1' .. ii',"! :.',:*,*:;I J 2. An immediate notification of the occurrence shall be made to the Reactor Supervisor, Reactor* Administrator, and ROC; and 34

3. A report, and any applicable follow-up report, shall be, prepared and submitted to the NRC. The report shall describe the following:

']*:*, ' *. . . .. . '. a. Applicable circumstances leading to the violation including, when known; the and contributing factors; b. Effects of the violation upon reactor facility components, systems; or structures arid on the health and safety of personnel and the public; and c. Corrective action to be taken' to recurrence:

6.6.2 Actions

to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 Other than a Safety Limit Violation . * ' . For all events which are required by Technical Specifications to be reported to the NRC within 24 hours ., ' . . ' ,*., .. under TS 6.7 .2, except a safety limit violation, the following actions.shall b*e taken: 1. The reactor shall be secured and the Reactor Supervisor notified; . -. . :* . '.1.: 2. Operations shall not resume unless authorized bythe Reactor Supervisor;

3. The ROC shall review the occurrence at their next scheduled meeting; and 4. Where appropriate, a report shall be submitted to the NRC in accordance with TS 6.7.2: 6.7 Reports 6. 7.1 Annual Operating Report * , : * , " 11* An annual report covering the previous catendar'year'shall be created and submitted, no later than* March 31 of the year following the report period, by the Reactor Supervisor to the NRC consisting'of:

1. A brief summary of operating experience including the energy produced by the reactor and the hours the reactor was critical;

2. The number of unplanned shutdowns, including corrective actions taken (when applicable);

3. A tabulation of major preventative and corrective maintenance operations having safety . significance;

4. A brief description, including a summary of the safety evaluations, of changes in or in procedures and of tests and experiments carried out pursuan.t ti;i :).0 CFR .50.59; ',*-, 5. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.

The summa*ry shall include to the ei<fent p'racticable 'ari estimate-of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25% of the allowed effect is sufficient; 35 .* ; * ,

6. A summarized result of the facility;

7. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed; and . . . * . .

• 8. Results of fuel inspections (when performed).

6.7.2 Special

Reports In addition to the requirements of applicable regulations, and in no way supst\tuting th.erefpre, reports shall be made by the Reactor Supervisor to the NRC as follows: 1. A report within 24 hours by telephone, confirmed by digital submission; or fax the NR,C ' ' Operations Center if requested, and followed by a report in writing to the NRC, Document Control Desk, Washington, 6.c.* V:.it,hin 14 with any of the following:

• . . . . . a. Any release of radioactivity above applicable lim'its into unrestricted

areas, or not the release resulted in property damage, personal injury, or exposure;

b. Any violation of a safety limit; c. Operation with the actual safety system setting less conservative than the LSSS; d. Operation in violation of a Limiting Condition for Operation;

e. Malfunction of a required reactor safety system component which renders or could** , ,. render the system incapable of performing its intentjed s,afety function.unless the. ,_; .>.*" fai4!fe-malfunction or condition is aiscovereEl Eluringcaused by maintenance-te-st£-e.F

,.,),* ,.1 :' perioEls of reactor shutElown, then no report is required;

f. Any unanticipated or uncontrolled change in reactivity greater than $1.00. Reactor trips resulting from a known cause are excluded;

g. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadeq'uacy causes or could have caused the existence or development of a condition which results or could result in operation of the reactor outside the specified safety limits; or '1* . -, . .** .1 *':1 h. Abnormal and significant degradation ih reactor fuel, cladCing, cir' coolant boundary !'.* ** ; . ; ... * '* f

'* l(J '. :1) 1 I , , '1*' 2. A report within 30 days in NRC:; D,ocuJTlel)t C:ontrol Desk, Washington,.D.C. of: a. Permanent in the Level 1-2 personnel; or *, .. '. ;.,-*** '* ' " . . , . ' 36

6.8 Records

b. Significant changes in the transient or accident analyses as described in the Safety Analysis Report. 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved ifLess than Five Years 1. Normal reactor operation

{but not including supporting documents such as checklists, data sheets, etc., which shall be maintained for a period of at least two years); 2. Principal maintenance activities;

3. Reportable occurrences;

4. Surveillance activities required by the Technical Specifications;

5. Reactor facility radiation and contamination surveys; 6. Experiments performed with the reactor; 7. Fuel inventories, receipts, and shipments;

8. Approved changes to the operating procedures; and 9. ROC meetings and audit reports. 6.8.2 Records to be Retained for at Least One Operator License Term 1. Records of retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one license term; and 2. Records of retraining and requalification of licensed operators shall be maintained while the individual is employed by the licensee, or until that operator's license is renewed, whichever is shorter. 6.8.3 Records to be Retained for the Lifetime of the Reactor Facility 1. Gaseous and liquid radioactive effluents released to the environs;

2. Offsite environmental monitoring surveys; 3. Reviews and reports pertaining to a violation of the safety limit, the limiting safety system setting, or a limiting condition of operation;

4. Radiation exposures for all personnel monitored; and 5. Drawings of the reactor facility.

37}}