ML20217K770

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Amend 12 to License R-97,removing Authority from License to Operate,Authorizing possession-only & Changing TS to Remove Operational Requirements of Reactor
ML20217K770
Person / Time
Site: Neely Research Reactor
Issue date: 04/02/1998
From: Mendonca M
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217K622 List:
References
R-97-A-012, R-97-A-12, NUDOCS 9804070309
Download: ML20217K770 (21)


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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555 4001

.....,o GEORGIA INSTITUTE OF TECHNOLOGY DOCKET NO. 50160 AMENDMENT TO FACILITY LICENSE Amendment No.12 License No. R 97

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to Facility License No. R-97 filed by the Georgia Institute of Technology (Georgia Tech or the licensee), dated August 7,1997, as supplemented on November 6,1997, and February 4,1998, cornplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as set forth in 10 CFR Chapter I; B. The facility will be possessed in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations of the Commission; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the regulations of the Commission and all applicable requirements have been satisfied; and F. Prior notice of this amendment was not required by 10 CFR 2.105 and publication of notice for this amendment is not required by 10 CFR 2.106.

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2. Accordingly, the license is amended by changes to the following paragraphs which are hereby amended to read as follows:

A. This license applies to the heavy water moderated, tank-type nuclear reactor facility (the facility) owned by the Georgia Institute of Technology. The facility is located on the licensee's campus in Atlanta, Georgia, and is described in the applications dated February 1,1960, March 11,1968, and subsequent amendments thereto.

i B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Georgia institute of Technology:

(1) Pursuant to Section 104c of the Act and 10 CFR Part 50 " Domestic Licensing of Production and Utilization Facilities," to possess, but not operate the reactor at the designated location in Atlanta, Georgia, in accordance with the procedures and limitations described in the application and this license.

(2) Pursuant to the Act and 10 CFR Part 30 " Rules of General Applicability to Domestic Licensing of Byproduct Material." to possess a 50-curie antimony-beryllium sealed neutron source.

(3) Pursuant to the Act and 10 CFR Part 30 " Rules of General Applicability to Domestic Licensing of Byproduct Material," to possess, but not to separate such byproduct material as may have been produced by past operation of the reactor.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, and Sections 50.54 and 50.59 of Part 50; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.12, are hereby incorporated in the license. The licensee g shall possess the facility in accordance with the Technical Specifications.

l (2) Physical Security Plan The licensee shall maintain and fully implement all provisions of the Commission's approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of documents withheld from public disclosure pursuant j

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&id 3 to 10 CFR '2.790, entitled " Proposed Pliysical Security Plan for Georgia T Research Reactor, Docket No. 50-160" dated June 10,1980, as revised by letters dated June 19,1981 and September 10,1981.

4.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Marvin M. Mendonca, Acting Director Non-Power Reactors and Decommissioning Project Directorate

. Division of Reactor Program Management Nuclear of Reactor Regulation

Enclosure:

Appendix A Technical Specifications Changes Date of issuance: April 2,1998 t

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'3j ENCLOSURE TO LICENSE AMENDMENT NO.12 FACILITY LICENSE NO. R-97 DOCKET NO. 50-160 Replace the Appendix A Technical Specifications in its entirety with the enclosed pages.

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9 GEORGIA INSTITUTE OF TECHNOLOGY NEELY NUCLEAR RESEAR.CH CENTER TECHNICAL SPECIFICATIONS TO OPERATING LICENSE NO. R-97 DOCKET NO. 50-160 REVISION TO FACILITATE DECOMMISSIONING e

TABLE OF CONTENTS 1.0 DE FINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 LIMITING 2.1 CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . 3 2.2 REACTIVITY LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 SAFETY SYSTEMS 2.3 ...................................

2.4 CONTAINMENT BUILDING . . . . . . . . . . . . . . . . . . 5. . . . . . .

RADIOACTIVE EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 3.2 SAFETY RELATED INSTRUMENTATION . . . . . . . . . . . . . . . . . . . 8 CONTAINMENT BUILDING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.0 SITE DESCRIITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.0 ADMINISTRATIVE CONTROIE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.1 5.2 ORG ANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 I 5.3 TECHNICAL SAFETY REVIEW COMMITTEE . . . . . . . . . . . . . . . 11 5.4 PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 {

5.5 OPERATING RECORDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 i ACTION TO BE TAKEN IN THE EVENT OF A REPOR 5.6 OCCURRENCE . . . . . . . . . . . . .. . . . . . . . . . . . . .14. . . . . . . . .

REPORTING REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 14 1

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l 1.0 DEFINITIONS 1.1 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification shall include comparison of the channel with other independent channels or methods measuring the same variable.

1.2 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within acceptable range and accuracy, to known values of the parameter '

which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip.

1.3 Containment Intecrity - Containment integrity exists when all of the following conditions are met:

a. One door on each personnel air lock is closed and scaled.
b. The tmck door is closed and sealed.
c. Controls, equipment and interlocks for isolation of the containment building are operable or the containment is isolated.

1.4 Decommissionine - 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 for unrestricted use and termination of the license; or (2) release of the property under restricted conditions and termination of the license.

1.5 E3 contamination - Decontamination are the activities employed to reduce the levels of radioactive and/or hazardous contamination in or on material, structures, and equipment.

1.6 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.

1.7 local Con!.amment - Use of an enclosure and procedures to prevent or mmunize the spread of contamination, often with HEPA-filtered ventilation.

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

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

1.10 Reactor Facility - Reactor facility means the containment structure located at the Neely Nuclear Research Center and everything contained inside the structure to include 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.

1.12 Reportable 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 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 limits specified in 10 CFR Part 20.
c. An ebserved inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with activities performed under the USNRC possession only license.

1.13 Surveillance Frecuency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations and exammations 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% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval.

1.14 Surveillance Interval - The surveillance interval is the calender time between surveillance tests, checks, calibrations and examinations to be performed 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, j component or system shall be tested prior to being declared operable.

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LIMITING CONDITIONS FOR OPERATION -

2.1 REACTIVITY LIMITS l APPLICABILITY I This specification applies to the reactivity condition of the reactor, j OBJECTIVE To assure that the reactor is shutdown at all times and that the safety limits will not be exceeded.

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

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BASIS The reactor cannot be operated without nuclear fuel.

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l 2.2 SAFETY SYSTEMS APPLICABILITY This specification applies to safety related instrumentation and systems associated with licensed activities.

OBJECTIVE 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 monitored to ensure that radiological releases to the environment are within acceptable levels as specified in 10 CFR Part 20. Table 2.1 provides a list of safety related instmmentation 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 i

Tritium Monitor (c) _ < 10 mR/hr 1(b) Alarm & initiate containment -

isolation -

Moving Filter <5 mR/hr 1(b) Alarm & initiate containment j Particulate Monitor isolation i Filter Bank Monitor <5 mR/hr 1(b) Alarm & initiate containment isolation j (a) Area monitors shall be located on the experimental level, the reactor top and in the reactor facility basement.

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

or calibration. 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 i from the reactor facility.

(d)

Set points will be based upon radiation levels in Table 2.1 converted to specific instrument readouts (

BASIS 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 other system would provide indication of an abnormal condition.

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. ,, i 2.3 CONTAINMENT BUILDING APPLICABILITY This specification applies to the reactor facility containment building requirements 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 To minimize the release of airborne 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). '

BASIS Building containment is a major engineered safety feature which served as the final 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 term is removed. 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 Preparing 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 operated. Once the heavy water is removed from the reactor facility, a major source of tritiurn is eliminated.

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

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3sl OBJECTIVE ,

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

a. Liauid Wastes (1) The concentration of gross radioactivity, above background, discharged from the reactor facility to the sanitary sewer during licensed activities 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 radioactivity in that sample is within the limits of 10 CFR Part 20, release to sanitary sewerage system may begin. The process of discharging the liq-.o 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 meet criteria for release following processing, as an option, the liquid can be treated and solidified for future disposal at a licensed waste disposal facility.

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

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

(2) During licensed activities involving possible production of airborne 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 the 10 CFR Part 20 limits are not exceeded.

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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 licensed 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.

BAS!S i

a. Liould Wastes

'Ibe liquid waste handling system is described in the Safety inalysis 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 determme 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 release rate limit for gross radioactivity takes into account local meteorological data. The release rate limit for radionuclides and particulates with halflives longer than eight days takes anmm1 average atira,.y gric 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 2

value of X/Q which is 4.1 x 10d sec/m , for the 22.5' sector having the least 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 abnormally high gross Page 7 of 15 l

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1 radioactivity releases due to licensed activities while heavy water (not to incluce residual holdup) is stored at the reactor facility.

3.0 SURVEILLANCE REQUIREMENTS i

3.1 SAFETY RELATED INSTRUMENTATION APPLICABILITY l

These specifications apply to the surveillance of safety related instrumentation.

OBJECTIVE To assure that the systems important to safety are operable as required. I 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 Semiannudly Source Calibration Annually Building Radiation Monitor x (1) x x Tritium Monitor x (2) x x Moving Filter Particulate x (2) x x Monnor Filter Bank Monitor x G) x x (1) Applicable only when licensed activities are being performed which couM cause a radiation area.

I (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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W BASIS 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 p escribed' input signals. The Georgia Tech Radiation Safety Manual specifies that derived air concentrations shaJ1 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 To verify containment building integrity and to determine and record the building' leakage rate under test conditions.

SPECIFICATION

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

- approximately six month intervals,

b. An integrated leakage rate test of the containment building shall be performed annually at a pressure of at least 2.0 psig. Imakage 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 additions, modifications or maintenance of the containment bt'ilding or its penetrations shall be tested to verify that the building can maintain its required leak tightness.

BASIS 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 Preparmg 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 reason for maintaining contamment 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 w+% .

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was the reactor itself. During the possession only licensed period, the reactor will not be operated.

Containment building isolation is initiated by a signal from the tritium monitor, the moving particulate monitor, the filter bank monitor or a manual push button on the reactor console. Operability and trip point checks semi-annually will be performed in accordance 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 trend has developed which would indicate a gradual deterioration of the containment building. An annual leak rate test frequency is therefore consistent with past experience.

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

4.0 SITE DESCRIPTION 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 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 reactor stack exhaust.
c. The exclusion area is the area inside the circle formed by a 100 meter (328 ft) radius centered at the reactor.
d. The low population zone outer boundary is formed by a 400 meter (1312 ft) radius from the reactor center.
c. The population center distance for the reactor facility is established as radius of 523 meters (1750 ft) from the reactor center.

5.0 ADMINISTRATIVE CONTROLS 5.1 ORGANIZATION Page 10 of 15

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The organization for the management and oversight for the reactor facility during the time when the possesion only license is in effect shall be as indicated in Figure 5.1. The Director, Neely Nuclear Research 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 exposure and adhering to all requirements of the possession only license and Technical Specifications.

Georgia Tech Organization for Management of the Reactor Facility President GeorgiaInstitute of Technology Dean TechnicalSafety Vice Provost for Research College of Engineering Review Committee & Dean of Graduate Studies i

Neely Nucl

," #7 Re ch Center Figure 5.1

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

5.2 TECHNICAL SAFETY REVIEW COMMITTEE

a. The Technical Safety Review Committee shall be established by the President of the Institute and shall be responsible for maintaining health and safety standards Page 11 of 15 t

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b. The Committee shall be composed of four or more senior technical personnel who collectively provide experience in reactor operations, radiochemistry, radiological safety, radiation protection, and mechanical and electrical systems.
c. The committee shall meet semiannually or as circumstances warrant. Written records of the proceedings, including any 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 Committee membership and shall include the chairman or his designated alternate.

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

(2) Review reportable occurrences.

(3) Review and approve proposed procedures and proposed changes involving licensed activities. Minor modifications to procedures which do not change .

the original intent of the procedures may be approved by the Director of the I NNRC on a temporary basis. The Committee shall consider for approval such minor modifications at its next scheduled meeting.

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

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

(6) Review violations of the Technical Specifications, license or internal l

procedures or instructions having safuy significance.

G) Review abnormalities occurring during licensed activities having safety significance.

(8) Review audit reports.

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

(10) Audit the results of action taken to correct those deficiencies that may occur la the reactor facility equipment, systems, structures or methods of operation that affect safety, at least once per calendar year (interval between audits not to exceed 15 months).

5.3 PROCEDURES

a. Licensed 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 Saferj- Review Committee prior to being effective. Changes which do not alter the original intent of a procedure may be cpproved by the Director of 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 operation of all systems related to safety.

(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 period 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 (1) Normal facility operation and maintenance.

(2) Reportable occurrences.

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

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

(5) Off-site environmental monitoring surveys."

(6) Facility radiation and contammation surveys.

(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 indermitely or until-the USNRC authorizes their disposal.

5.5 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE In the event of a reportable occurrence, as defined in these Technical Specificatiorc,, the following action shall be taken:

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a. Ongoing activities shall be ceased until the occurrence has been resolved.
b. l All reportable occurrences shall be promptly reported to the Director of the Neely Nuclear Research Center or designee.
c. All reportable occurrences shall be reported to the USNRC in accordance with appropriate regulations,
d. All reportable occurrences shall be reviewed by the Technical Safety Review Committee.

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

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

a. Reportable Occurrence Report Notification shall be made ivithin 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 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 the 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 measures 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 30 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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