ML20202F462

Proposed Tech Specs Re Rev to Facilitate Decommissioning

ML20202F462
Person / Time
Site:	Neely Research Reactor
Issue date:	02/04/1998
From:	Neely Research Reactor, ATLANTA, GA
To:
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ML20202F451	List: ML20202F447 ML20202F462
References
NUDOCS 9802190205
Download: ML20202F462 (17)

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GEORGIA INSTITUTE OF TECHNOLOGY NEELY NUCLEAR RESEARCH CENTER TECHNICAL SPECIFICATIONS TO OPERATING LICENSE NO. R-97 r

DOCKET NO. 50-160 REVISION TO FA.CILITATE DECOMMISSIONING Revision - 2/4/98

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4 TABLE OF CONTENTS 1.0 DE FINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . 3 2.1 REACTJVITY LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 SAFETY SYSTEMS ................................... 3 2.3 CONTAINMENT BUILDING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.4 RADIOACTIVE EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 SURVEILLANCE REQUIRESENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 SAFETY RELATED INSTRUMENTATION . . . . . . . . . . . . . . . . . . . 8 3.2 CONTAINMENT BUILDING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

• 4.0 SITE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.1 OR G ANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.2 TECHNICAL SAFETY REVIEW COMMITTEE . . . . . . . . . . . . . . . 11 5.3 PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5.4 OPERATING RECORDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5.5 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OC C URREN C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 5.6 REPORTING REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 14 1

1 4-h 1.0 DEFINITIONS I l

1.1 Channel Check - A channel check is a qualitative verification of acceptable performance i .by observation of channel behavior. This verification shall include comparison of the channel with other independent channels or methods measuring the same variable.

l.2 Channel Calibration - A channel calibration is a- . .t of the channel such that its

_ output responds, within acceptable range and accura c wn values of the parameter which the channel measures. Calibration shall encony a die entire channel, including equipment actuation, alarm or trip.

- 1.3 Containrent Integrity . Containment integrity exists when all of the following conditions are met

' a. One door on each personnel air lock is closed and sealed,

b. The truck door is closed and scaled.

c. Controls, equipment and interlocks for isolation of the containment building are operable or the containment is isolated.

1.4 - Decommissionino - Decommissioning means to remove a facility or site safely from

, service and reduce residual radioactivity to a level that permits: (1) release of the property -

I' for unrestricted use and termination of the license; or (2) release of the property under restncted conditions and termination of the license.

4 1.5 Decontamination - Decontamination are the activities employed to reduce the levels of p radioactive and/or hazardous contamination in or on material, structures, and equipment.

l.6 Licensed Activity - Licensed activity means any activity or operation authorized to be performed during the period of time when a possession only license is in force for the reactor facility. Typical activities that may be performed during this period include, but are not limited to, packaging and shipment of waste generated from previous GTRR operations and from - possession only-. activity reactor facility waste, normal decontamination of components, surveillance and maintenance of equipment, and decommissioning. Reactor operation is not an activity which can be performed during this time. Even though decommissioning is an activity that can be performed during this

• period, it is understood that decommissioning activities can not be performed without prior NRC acceptance of the decommissioning plan.

1 17 Local Containment - Use of an enclosure and procedures to prevent or minimize the spread of contamination, often with HEPA-filtered ventilation.

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1.8 Operable - Operable sneans a component or system is capable of performing its intended function in its normal manner.

1.9. .Oneratina - Operating means a component or system is performing its intinded function in its normal manner.

1.10 Reactor Facility - Reactor facility means the conti anent structure located t the Neely Nuclear Research Center and everything contained inside the structure to ielude the reactor, support systems and remaining equipment. It does not imply operation of the reactor as a neutron or other energy source.

1.11 - Reactor Shutdown - Reactor shutdown means that there is no fuel in the core.

l.12 Renortable Occurrence - A reportable occurrence is any of the following:

. a. An uncontrolled or unplanned release of radioactive material from the restricted area of the reactor facility which results in a concentration that exceeds the limits j set forth in 10 CFR Part 20,

b. An uncontrolled or unplanned release of radioactive material which results in concentrations of radioactive materials within the reactor facility in excess of the l limits specified in 10 CFR Part 20. ,

c. An observed inadequacy in the unplementation of administrative or procedural ,

controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe cond' tion in connection with activities performed u> der the USNRC possession only license. -

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1.13 Surveillance Freauency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus or minus 25%. In cases where the elapsed interval has exceeded 100% cf the specified interval, the next surveillance laterval shall commence at the end of the original specified interval.

1.14 Survei11 mace Interval - The surveillance interval is the calender time between surveillance tests, checks, calibrations end examinations to be perform:d upon any instrument or component when it is required to be Operable. These tests may be waived'when the

-instrument, component or system is not required to be operable, but the instrument, component or system shall be tested prior to being declared operable.

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4 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 REACTIVITY LIMITS APPLICABILITY -

This specification applies to the reactivity condition of the reactor.

OBJECTIVE To assure that the reactor is shutdort at all times and that the safety limits will not be exceeded.

SPECIFICATION All nuclear fuel remaina removed from the reactor and containment building. No fuel shall be reintroduced into these areas. -

BASIS The reactor cannot be opereed without nuclear fuel.

2.2 SAFETY SYSTEMS APPLICABIMTX

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l l This specification applies to safety related instrumentation and systems associated with licensed activities.

QBJECTIVE To specify the lowest acceptable level of performance and the minimum number of acceptable components for the safety related instrumentation and systems as determined by the Technical Safety Review Committee.

SPECIFICATION The reactor shall not be made critical. The safety related instrumentation and systems shall be monito: :d to ensure that radiological releases to the enviionment are within acceptable levels as specified in 10 CFR Part 20. Table 2.1 provides a list of safety related instrumentation that will be maintained during the licensed period when licensed activities related to radioactive materials are being performed. No license activities related to Page 3 of 15

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radioactive material will be performed unless the monitors specified in Table 2.1 are operable.

TABLE 2.1 SAFETY RELATED INSTRUMENTATION REQUIRED FOR LICENSED ACTIVITIES Instrumentation Set Point (d) Minimum Function Number Required Building Radiation < 5 mR/hr 3(a) Alarm Monitor Tritium Monitor (c) < 10 mR/hr 1(b) Alarm & initiate containment isolation Moving Filter < 5 mR/hr 1(b) Alarm & initiate containment Particulate Monitor isolation Filter Bank Monitor < 5 mR/hr 1 (b) Alarm & initiate containment isolation (a) Area monitors shall be located on the experimental level, the reactor top and in the reactor factity basement.

, (b) Ar instrument may be inoperable for a period not to exceed 3 working days for test, repair l

or il%iion. No more than one of these systems may be inoperable at any one time.

(c) Will remain in service until the heavy water (not including residual holdup) is removed

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from the reactor facility.

(d) Set points will he based upon radiation levels in Table 2.1 converted to specific instrument readouts M Sl3 Radiation monitoring will assure that areas throughout the reactor facility in which radiation areas could occur if conditions change during licensed activities, such as improper sample handling, equipment removal or shielding movements, are identified.

The filter bank monitor and the moving air particulate monitor provide redundant channels which monitor releases from the reactor facility. If one system is inoperable, the ot' er system would provide indication of an abnormal condition.

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2.3 CONTAINMENT BUILDING APPLICABILITY This specification applies to the reactor facility containment building requirements as long as heavy water (not including residual hnidup) is stored within the reactor facility. Once this ms,erial is removed from the reactor facility, this technical specification no longer applies.

OBJECTIVE To minimize the release of airl,orne radioactive materials from the reactor facility.

SPECIFICATION Containment integrity shall be maintained during licensed activities while heavy water remains on-site (except for residual holdup in piping).

EASIS Building containment is a major engineered safety feature which served as the imal physical barrier to contain radioactive particles and gases present or produced during reactor operations. Once the heavy water is removed from the site, a significant source l term is remove .. Reactor fuel has already been removed from the site. Local control will be used within the building when necessary to provide an additional contamination control boundary.

NUREG-1537, Part 2, " Guidelines for Prepar.ng and Reviewing Applications for Licensing of Non-Power Reactors", identifies the main purpose of a containment is to mitigate accident consequences for the operating reactor. The major source of potential -

over pressurization during the operating period of the reactor was the reactor itself. During the possession only licensed period, the reactor will not be opertted. Once the heavy water is removed from the reactor facility, a major source of tritium is eliminated.

- 2.4 RADIOACTIVE EFI1UENTS APPLICABILITY This specification applies to the controlled release of radioactive liquids and gases from the reactor facility.

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, QEIECTIVE To define the limits and conditions for the release of radioactive effluents to the environs to' assure that any radioactive effluents are as low as practical and would not result in radiation exposures greater than a few percent of natural background exposures, and within the limits of 10 CFR Part 20.

SPECIFICATIONS 1

a. Liauld Wastes l- (1) The concentration of gross radioactivity, above background, discharged from the reactor facility to the sanitary sewer during licensed acth ities shall not exceed the limits of controlled discharge on a radionuclide basis set .

forth in 10 CFR Part 20 Appendix B, Table 3 and Notes.

(2) Before discharging'any liquid waste from any of the holdup tanks, the.

following shall be performed:

1. ' Isolate the tank to be emptied so that no liquid waste can be added during discharge.

2. Obtain a sample and analyze for content of radioactivity. If the radiactivity in that sample is within the limits of 10 CFR Part 20, release to sanitary sewerage system may begin. The process of.

discharging the liquid waste from the tank shall be stopped at a point past the halfway mark, but before the tank is 75 % discharged, to analyze another sample. If results of the second analysis are within specified limits, release of the rest of the liquid waste may-resume.

3. Liquid waste that fails to meet radioactivity release limits during decontamination activities will be processed and again sampled for possible release to the sanitary sewer. If results are within specified limits, as above, release of the liquid waste may proceed.

If- sample results fail to men. criteria for release following processing, as an option, the liquKl can be treated and solidified for future disposal at 1 licensed waste disposal facility.

b. Gaseous Effluents (1) The maximum release rates of gross radioactivity in gaseous effluents Page 6 of 15

. . during licensed activities shall not exceed the limlis set forth in 10'CFR Part 20 Appendix B Table 2, Column 1 and Notes.

(2)- During licensed activities involving possible production of' airborne 4

radioactivity, containment integrity shall be maintained and continuous air sampling will take place in an attempt to quantify the radiation exposure to personnel and to ensure that tl.e 10 CFR Part 20 limits are not exceeded.

c. Liquid Effluents (1) The maximum release rates of gross radioactivity in liquid effluents during '

licensed activities shall not exceed the limits set forth in 10 CFR Part 20, Appendix B, Table 2, Column 2 and Notes.

(2) During licensM activities . involving possible production of liquid radioactivity, all liquid waste will be directed to holding tanks where surveys shall be performed to ensure that the 10 CFR Part 20 limits are not exceeded.

BASIS

a. Liauld Wasteji

- The liquid waste handling system is described in the Safety Analysis Report dated January 10,1995. The maximum amount of tritium in the discharge is limited to -

the value given in 10 CFR Part 20. The total quantity of radioactivity limit is in accordance with 10 CFR Part 20 for disposal to a sewage system. The independent samples taken prior to and during liquid effluent release shall determine the radioactivity -concentration in the liquid released from the tanks and the radioactivity concentration in the discharge line to the sanitary sewers.

- b. Gaseous Effluents The relecse rate limit for gross radioactivity takes into account local meteorological data. The release rate limit for radionuclides and particulates with half lives longer than eight days takes annual average atmospheric dilution into account and ensures -

that at any point on or beyond the restricted area fence the requirements of 10 CFR Part 20 will be met. The limit is based on the annual average diffusion coefficient value of X/Q which is 4.1 x 10d sec/m', for the 22.5* sector having the least

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diffusion on an annual average.

Isolation of the exhaust effluent stack is initiated by high radiation in the off-gas system. Such isolation is required- for potentially abnornully high gross Page 7 of 15

radioactivity releases due to licensed activities while heavy water (not to include residual .hok cp) is stored at the reactor facility.

3.0 SURVEILLANCE REQUIREMENTS 3.1 SAFETY RELATED INSTRUMENTATION l APPLICABILITY ,

These specifications apply to the surveillance of safety related instrumentation.

OBJECTIVE To assure that the systems important to safety are operable as required.

SPECIFICATIONS The channels listed in Table 3.1 shall be checked and calibrated as indicated.

TABLE 3.1 SURVEILLANCE REQUIREMENTS FOR SAFETY RELATED INSTRUMENTATION Surveillance Requirements '

Channel Daily Source Check Known Parameter Check Semiannually Source Calibration Annually Building Radiation Monitor x (1) x x Tritium Monitor x (2) x x Moving Filter Particulate x (2) x x Monitor Filter Bank Monitor x (2) x x (1) Applicable only when licens"3 activities are being performed which could cause a radiation area.-

(2) Applicable only when licensed activities are being performed which could cause a potential airborne release to the containment, i.e., greater than 12 DAC-hours per week.

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BASIS 4

Calibration of the safety related instrumentation means to measure the performance as guided by the vendors instructions and performance specifications of the instrument in its .

response to accurately prescribed input signals. The Georgia Tech Radiation Safety Manual specifies that derived air concentrations shall be monitored if an individual, without respiratory protection, could potentially exceed 12 DAC-hours per week.

3.2 CONTAINMENT BUILDING-APPLICABILITY This specification applies to the surveillance of the containment building as long as heavy water (not including residual holdup) is stored within the reactor facility. Once this material is removed from the reactor facility, this technical specification no longer applies.

OBJECTIVE I

To verify containment building integrity and to determine and record the building leakage rate under test conditions.

l SPECIFICATION -

a. The containment building isolation initiating system shall be tested twice a year at.

appro? untely six toonth intervals.

. b. - An integrated leakage rate test of the containment building shall be performed annually at a pressure of at least 2.0 psig. Irakage from the building'shall not exceed 1.0% of the building air volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> overpressure,

c. All addhions, modifications or maintenance of the containment building or its penetrations shall be tested to verify that the building can maintain its reauired leak tightness.

RASIS There is minimal possibility for the containment building pressure to exceed 2.0 psig during any postulated activities associated with licensed activities performed during the possession only licensed period. NUREG-1537, Part 2, " Guidelines for Preparing and Reviewing Applications for Licensing of Non-Power Reactors", identifies the maic purpose of a containment is to mitigate accident consequences for the operating reactor.

The reason for maintaining containment integrity during the possession only licensed period is to mitigate any releases to the environment from a spill of the stored heavy water. The major source of over pressurization during the operating period of the reactor Page 9 of 15

4 was the ret.: tor itself. During the possession only licensed period, the reactor will not be operated.

Comainment building isolation is initiated by a signal from the tritium monitor, the moving >

partkulate monitor, the filter bank monitor or a manual push button on the reactor console. Operability anJ trip point checks semi annually will ba performed in accordarme with Table 3.1.

The containment has been leak tested annually since 1%3 and the leak rate of 1.0% of the building volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has never been exceeded. No trer) has developed which would indicate a gradual deterioration of the conte.inment buuding. An annual leak rate test frequency is therefore consistent with past experience.

Any additions, modificati;ns or maintenance to the buildmg or its penetrations will be tested to verify that such work has not adversely affected the leak tightness of the building.

4.0 SITE DESCRIIrliON SPECIFICATION

a. The reactor facility is located on the Georgia Institute of Technology campus in the city of Atlanta, Georgia,

b. The restricted area is formed by eight foot perimeter fence on the east, south and 1 west of the containment building and the laboratory building on the north. The closest unrestricted area is 40 meters in a West Southwest direction from the l reactor stack exhaust.

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c. The exclusion area is the area inside the circle formed by a 100 meter (328 ft) radius centered at the reactor, t d. The low population zone outer boundary is formed by a J00 meter (1312 ft) radius I from the reactor center,

e. The population center d: .nce for the reactor facility is established as radius of 523 metcrs (1750 ftp from the reactor center.

5.0 ADMINISTRATIVE CONTROLS 5.1 ORGANIZATION Page 10 of 15 l

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a. The organization for the management and oversight for the reactor facility during

! .- tt , time whe1 the possession only license is in effect shall be as indicated in Figure 5.1. The Director, Neely Nuclear Researcia Center (NNRC) shall have overall responsibility for direction and performance of activities at the reactor facility.

including safeguarding the general public and facility personnel from radiation

' exporure and adhering to all requirements of the possession only license and Technical Specifications.

Georgia Tech Organization for Management of the Reactor Facility 1

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President l GeorgiaInstitute of Technology ,

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I Dean - Technical Safety Vice Provost for Research College of Engineenng Review Committec & Dean of Graduate Studies 4

NeelyNucl h es h Center h"c,f**Y 3-Figure 5.1 5

b. The Radiation Safety Officer shall advise the Director, Neely Nuclear Research Center in matters pertaining te radiological safety. He/she has access to the Vice-2, Provost, Research and Graduate Studies and/or the Pre:Ident of the Institute as needed.

5.2 - TECHNICAL SAFETY REVIEW COMMITTEE

a. The Techulnl Safety Review Committee shall be established by the President of the institute and shall be responsible for malt.taining health and safety standards Page !I of 15

l associated with the licensed activities during the possession only licensed period

. for the reactor facility,

b. ' The Committee shall be composed of four or more senior technical personnel who l collectively provide ex;erience in reactor operations, radiochemistry, radiological safety, radiation protection, and mechanical and electrical systems,

c. The committee shall meet semiacnuaily or as circumstances warrant. Written records of the proceedings, including coy recommendations or occurrences, shall be distributed to all Committee members and the President's Office.

c. The quorum shall consist of not less than a majority of the Conunittee membership and shall include the chairman or his designated alternate,

d. *lhe Committee shall:

(1) Review and approve determinations that proposed changes in equipment, systems, tests, or proccJures do not involve an unreviewed safety question pursuant to 10 CFR 50.59 (a).

-(2) -- Review reportable occurrences.

(3) 11eview and approve proposed procedures and proposed changes involving l licensed activities. Minor modifications to procedures which do not change the original intent of the procedures may be approved by the Director of the NNRC on a temporary basis. The Committee shall consider for approval such minor modifications at hs next scheduled meeting.

(4)- Review and approve proposed changes to the Technical Specifications and license excluding organizational structure. The responsibility and authcity for organizational structure resides with the Fresident of the Institute.

(5) Review and approve proposed changes to the reactor feility made pursuant to 10 CFR 50.59 (a),

(6) - Review violations of the Technical Specifications, license ut internal procedures or instructions having safety significance.

(7) Review abnormalities occurring during licensed activities basing safety-significance.

-(8) Review audit reports.-

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(9) Audit records maintained during the Nensed period for compliatice with internal rules, procedures, and regulations and with licensed provisions including TechetA Specifications at least once per calendar year (interval between audits not to exceed 15 months).

(10) Audit tha results of action taken to correct those deficiene.les that nay occur in the reactor fact!ity equipment, systems, structures or methods of operation that affect safety, at least once per calendar year (inten'al between audits not to exceed 15 months).

5.3 PROCEDURES

a. 8 icensed activities shall be performed in accordance with approved written procedures. All procedures for licensed activities and major changes thereto shall be reviewed and approved by the Technical Safety Review Committee prior to being effective. Changes which d< 'mt alter the original intent of a procedure may be approved by the Director cf the facility. Such changes shall be recorded and

, submitted periodically to the Technical Safety Review Committee for routine review.

b. Written procedures shall be provided and utilized for the following:

(1) Normal opervion of all systems related to safety.

l (2) Actions to be taken in response to alarms.

(3) Emergency conditions involving potential or actual release of radioactivity.

(4) Radiation and radioactive contamination control.

(5) Surveillance and testing procedures.

(6) Physical security of the facility.

5.4 OPERATING RECORDS

a. in addition- to the requirements of applicable regulations, and in no way substituting therefore, records of the following items shall be maintained for a pericd of at least five years or as long as the facility is licensed by the U.S.

Nuclear Regulatory Commission (USNRC) for possession of radioactive material:

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(1) Normal facility operation and mainienance.

(2) Reportable occurrences.

(3) Tests, checks and measus ements documenting compliance with surveillance requirements.

(4) Gaseous and liquid waste released to the environment.

(5) Off sac environmental monitoring swveys.'

(6) Facility radiation and contamination sut veys.

(7) Updated, corrected and as built facility drawings.

(8) Minutes of the Technical Safety Review Committee meetings.

(9) Records of radioactive shipments,

b. Records of radiation exposure for all personnel shall be kept indefinitely or until the USNRC authorizes their disposal.

5.5 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE i

la the event of a reportable occurrence, as defined in these Technical Specifications, the following action shall be taken:

a. Ongoing activities shall be ceased until the occurrence has been resolved.

b. All reportable occurrences shall be promptly reported to the Director of the Neely Nuclear Research Center or designee, i

c. All reportable occurrences shall be reported to the USNRC in accordance with r.ppropriate regulations.

d. All reportable occurrences shall be reviewed by the Technical Safety Review Committee.

5.6 REPOR11NG REQUIREMENTS In addition to the requirements of applicable regulations, and in no way substituting Page 14 of 15 4

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- 0 therefore, reports shall be made to the USNRC as follows:

a. Pmrtable Ocmerene. Danart Notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone to the NRC's Operation F

Center, to be followed by a written report within 14 days to the U.S. Nuclear -

Regulatory Commission, Document Control Desk, Washington, DC, 20555 in ae event of a reportable occurrence as defined in Section 1.0. The written report on these reportable occurrences, and to the extent possible, the preliminary telephone call shall:

(1) describe, analyze and evaluate safety implications; (2) outline the awasures taken to assure that the cause of the condition is determined; (3) indicate the corrective action (including any changes made to the procedures and to the quality assurance program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems; and (4) evaluate the safety implications of the incident in light of the cumulative experience obtained from the record of previous failures and malfunctions of similar systems and components,

b. Unusual Events A written report shall be forwarded within 39 days to the Director, Office of Nuclear Reactor Regulation with a copy to the Office of the Regional Administrator, Region II in the event of discovery of.any unusual event,- i.e; substantial variance from performance specifications contained in the Technical Specifications.

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