ML20198T437

Submits Addl Info in Support of possession-only-license Amend & in Response to 970916 RAI

ML20198T437
Person / Time
Site:	Neely Research Reactor
Issue date:	11/06/1997
From:	Davidson J Neely Research Reactor, ATLANTA, GA
To:	Mendonca M NRC (Affiliation Not Assigned)
References
NUDOCS 9711140286
Download: ML20198T437 (37)

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@{ ceATbchmx A w Collegeof Engineering Office of the Dean November 6,1997 Docket No. 50-160 United States Nuclear Regulatory Commission l 1155$ Rockville Pike j Rockville,MD 20852 2738 Attentioc: Mr. Marvin M. Mendonca. Senior Project Manager i Non-Power Reactors and Decommissioning Project Directorce i Division of Reactor Program Management  ;

Omce of Nuclear Reactor Regulation -

Subject:

Possession-Only-License Amendment Application. AdditionalInformation

References:

1. USNRC Letter,

Subject:

Request for Additional information (TAC NO. M99370),

dated September 16,1997. . . .

2. Submittal: Possession Only License Amendment Request for Georgia institute of Technology Research Reactor, dated August 7,1997

- 3. Georgia Institute of Technology Rerah Reactor (GTRR) USNRC License No. R-97

4. NRC Docket No. 50-160

Dear Mr. Mendonca:

Georgia Tech is hereby submitting additional information in support of our Possession-Only License Amendment and in response to your request per Reference 1.

We would appreciate your timely review of this submittal and the processing of our Amendment request.

If you have any questions regarding this submittal, please contact our project manager, Mr. Bill Miller, at (404) 894 4610.

. Sincerely,

,)$

//

. Narl Davidson Associate Dean Chair, Technical Safety Review Committee Enclosures (2) as stated O' I

' 'c: (Enclosures Withdrawn): Dr. O. Wayne Clough. President Technical Safety Review Committec O

Collegeof Engineering Atlanta.Gwrgia 30332 0360 U.SA.

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GEORGIA INSTITUTE OF TECHNOLOGY

. NEELY NUCLEAR RESEARCH l CENTER ,

l TECHNICAL SPECIFICATIONS TO i 1

! OPERATING LICENSE NO. R-97 l

l DOCKET NO. 50-160  :

.I REVISION TO FACILITATE DECOMMISSIONING 4

i Enclosure 1 l

TABLE OF CONTENTS l

1 Pane I l

i 1.0 D E FI NI TI ONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ,

i 2.0 LIMITIWG CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . 3 2.1 REACTIVITY LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 SAFETY SYSTEMS ................................... 3 2.3 CONTAINMENT BUILDING . . . . . . . . . . . . ................ 4 2.4 RADIO /.CTIVE EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . P, 3.1 SAFETY RELATED INSTRUMENTATis N . . . . . . . . . . . . . . . . . . . 8 3.2 CONTAINMENT BUILDING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.0 SITE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.1 ORG ANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.2 TECHNICAL SAFETY REVIEW COMMITTEE . . . . . . . . . . . . . . . 11 5.3 PROC EDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5.4 OPERATING RECORDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5.5 ACTION TO BE TAKEN IN TIIE EVENT OF A REPORTABLE OCCU RR ENCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 5.6 REPORTING REQUIREMENTS . . . . . . . . . . . . . . . . . , . . . . . . . . 14 s

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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 Integrity - 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 truck door is closed and scaled.

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

1.4 Decommissioning - 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 Octontamination - Decontamin.. ion are the activities employed to reduce the levels of radioactive and/or hazardous contamination in or on material, structures, and equipment.

1.6 1.icensed 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 l decommissioning. Reactor operation is not an activity which can be performed during this l time. Even though decommissioning is an activity that can be performed during this l period, it is understood that decommissioning activities can not be performed without prior NRC acceptance of the decommissioning plan.

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1 j 1.7 local Containment - Use of an enclosure and procedures to prevent or minimize the spread of contamination, often with IIEPA-filtered ventilation.

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

1.9 Operating - Operati.ig means a component or system is performing its intended function

.n its nonnal mar.ner.

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 remr.ining 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 Reporr-bie Occurrengs - 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 observed 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 Survgillance Freauency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations and examinations shall be performed within the specilled 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, component or system shall be tested prior to being declared operable.

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t 2.0 - LIMITING CONDITIONS FOR OPERATION 2.1 REACTIVITY LIMITS APPLICABILITX This specification applies to the reactivhy condition of the reactor.

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 be scintroduced into these areas.

BASIS The reactor can not be operated without nuclear fuel.

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 or the minimum numbar 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 instrumentation that will be maintained during the licensed period.

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l TABLE 2.1 SAFLTY RELATED INSTRUh1ENTAT10N REQUIRED FOR LICENSED ACTIVITIES Instrumentation Set Point Minimum Function Number Required Building Radiation < 10 mR/hr or 3 (a) Alarm hionitor 2x background Tritium hionitor (c) Based on annual 1(b) Alarm & initiate containment calibration isolation Moving filter Based on annual 1(b) Alarn: & initiate containment par:iculate calibration isolation Filter bank monitor Based on annual 1(b) Alarm & initiate containment l calibration isolation (a) Area monitors shall be located on the experimental level, the reactor top and in the reactor facility basement.

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

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

3 ASIS Radiation monitoring will assure that areas throughout the reactor facility in which high 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, tritium monitor and the moving air particulate monitor provide diverse and redundant channels which monitor particulate and gaseous releases from the reactor facility.

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 Page 4 of 15

applies.

OBJECTIVE To minimize the release of airborne radioactive materials from the reactor facility.

SPECIFICATION Containment integrity shall be maintained during licensed activities which could cause airborne contamination levels to exceed 10CFR20, Appendix B, Table 2, Column 1,

" Effluent Concentrations".

IlAll3 Building containment is a major engineered safety feature which serves as the final physical barrier to contain radioactive particles and gases present or produced during licensed activities. Local control will be used within the bui! ding 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 reason for maintaining containment integrity during the possession only licensed period is to maintain the ALARA concept for off-site releases. The major sour ce of over pressurization during the operating period of the reactor was the reactor itself. During the possession only hcensed period, the reactor will not be operated. Once the heasy water is removed from the reactor facility, a major source of tritium is eliminated.

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

OBJECTIVF, 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.

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SPECIFICATIONS

a. Idauid Wastes (1) The cor. centration 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 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 aste 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 Effluents (1) The maximum release rates of gross radioactivity in gaseous effluents during licensed activities shall not exceed the limits set forth in 10 CFR Part 20 Appendix B, Table 2, Column I 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 Page 6 of 15

1 personnel and to ensure that the 10 CFR Part 20 limits are not exceeded.

c. Liauid 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.

BASIS

a. Liquid Wastes The liquid waste handling system is described in the Safety Analysis Report dated January 10,1995. The maximum vnount 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 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 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 3, 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 radioactivity releases due to licensed activities while heavy water (not to include residual holdup) is stored at the reactor facility.

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3.0 SURVEILLANCE REQUIREMENTS 3.1 SAFETY RELATED INSTRUMENTATION APPLICABILITY These specifications apply to the surveillance of safety related instrumentation.

QJUECTIVE 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 REQUIREhiENTS FOR SAFETY RELATED INSTRUhiENTATION Surveillance Requirements Channel Daily Source Check Known Parameter Check Semiannually Source Calibration Annually lluilding Radiation hionitor x (1) x x Tritium monitor x (2) x x hioving filter particulate x (2) x x Filter bank monitor x (2) x x (1) Applicable only when licensed activities are being performed which could cause a high radiation area.

(2) Applicable only when licensed activities are being performed which could cause an airborne release to the containment.

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 Page 8 of 15

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response to accurately prescribed input signals.

3.2- CONTAINMENT BUILDING APPLICABILITY This specification applies to the surveillance of the contaimnent building as long as heavy water (not including residual ho'iup) 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. Leakage 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 buildhig or its penetrations shall be tested to verify that the building can maintain its required leak tightness.

BASIS

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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 main purpose of a containment is to mitigate accident consequences fo:: 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 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 either the tritium monitor, the moving particulate monitor or a manual push button on the reactor console. Operabili:y and trip point checks semi-annually will be performed in accordance with Table 3.1.

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The containment has been leaked tested annually since 1%3 and the leak rate of 0.5% 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, ny 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,

e. 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

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

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Georgia Tech Organization for Manasent of the Reactor Facility Figure 5.1 President Georgia Institute of Technology Dean Technical Safety Vice Provost for Research College of Engineering Review Committee & Dean of Graduate Studies Director , , , , , , , , , ]. . . .. . . Radiation Safety Neely Nuclear Research Center Officer

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 TECIINICAL 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 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 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 Conunittee members and the President's Office.

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c. The quorum shall consist of not less than a majority of the Committee membership -l and shall include the chairman or his designated alternate, i d .' The C$mmittee shall: N (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. 1 i

(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 m:y be approved by the Director of the >

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 strr ture. 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 procedures or instructions having safety significance.

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

(8) Review audit reports.

(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 in 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).

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O 5.3 PROCEDURES

a. Licensed activities shall be performed in accordance with at. proved 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 do not alter the original intent of a procedure may be approved by the Director of the facility. Such changes shall be recorded and submitted periodical ly to the Technical Safety Review Conunittee 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.

(7) Preventive and corrective maintenance of systems related to safety.

5.4 OPERATING RECORDS

a. In addition to the requirements of applicable regulations, and m 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:

(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.

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_. . j (5) Off-site environmental monitoring surveys.

(6)- Facility radiation and contamination 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 indefinitely 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 Specificatioas, 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.

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

a. Reportable Occurrence Report 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 ar.d fax to the NRC's Operation Center, to be fe,. lowed 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 Page 14 of 15

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preliminary telephone call and fax notification shall:--

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(1)- describe, analyze and evaluate saRc/ 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 t

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 11 in the event of discovery of any unusual event, i.e.

substantial variance from performance specifications contained in the Technical' Specifications.

End of Document E

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GEORGIA INSTITUTE OF  !

TECHNOLOGY  !

XEELY NUCLEAR RESEARCH CENTER TECHNICAL SPECIFICATIONS

, TO OPERATING LICENSE NO. R-97 DOCKET NO. 50-160 REVISION TO FACILITATE DECOMMISSIONING Enclosure 2

9 TABLE OF CO: TENTS E482 1.0 . DEFINITIONS 1 2.0 LIMITING CONDITIONS FOR OPERATION 3 2.1 REACTIVITY LIMITS 3 2.2 SAFETY SYSTEMS- 3 2.3  ; CONTAINMENT BUILDING 4 2.4 RADIOACTIVE EFFLUENTS 5 3.0 SURVEILLANCE REQUIREMENTS 8 3.1 SAFETY RELATED INSTRUMENTATION 8 3.2 CONTAINMENT BUIL. DING 9 4.0 SITE DESCRII"I"ON ; 10 5.0 ADMINISTRATIVE CONTROLS 10 5.1 ORGANIZATION 10 5.2 TECHNICAL SAFETY REVIEW COMMITTEE 11 5.3 PROCEDURES 13 i- 5.4 OPERATING ~ RECORDS 13 5.5 ACTION- TO ' BE TAKEN IN THE EVENT OF A'. REPORTABLE OCCURRENCE .

14 l 5.6 REPORTING REQUIREMENTS ' 14 l

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1.0- 1 DEFINITIONS 1.11 - Chagpel 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 Cahoration -' 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. Calibraticn shall encompass the entire channel, inc.luding  !

equipment actuation, alarm or trip. '

1.3 Cherectcrizetien - Cherectcrization is the ; a n:inz. rnenitorina and 2n !vsis activities-to dctcrin'ac the cxtent and naturcContainman' fa'=itv iContainment integrity exists when -

all of contamination-the following conditions are met; iProvidesthe besis for acqairing thc necessarrtechnical information te pianihc accenwnissioning' activities:

h4 & itainment intcgrity - Containment 4ntegrity-exists whcn all-of the-following conditions are-mett

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

b. The truck door is closed and sealed,

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

h51,4 Decommissioning - Decommissioning is the action of-scaevingmeans to remove a facility or site safely from service and scducingreduce residual radioactivity to a level that permits:

(1) release of the property for unrestricted use and termination of the licenset or (2) release of the property under restricted conditions and termination of the license.

i h61.5 Decontamination - 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 Activitvl ; Licensed activity meansiany activity or operation authorized to be performed during .the_ period of time when al possession only license is in force for the reactor facilityETypical activities that may be7 performed _during this period include, but are not limitedito, packaging and: shipment of; waste l generated from previous GTRR operations:f andLfrom ; possession : onlyfactivity; reactorg facility; _ waste, : normal

- decontamination;of:, components,-(surveillance and; maintenance of equipment, and decommissioning. Reactor oleration is not an activity which can be performed during this Page 1 of 17

time. Even,though. decommissioning is an activity that can be performed durinF this perid it is understood that decommissioning activities can not be performed without prior NRC acceptance of the decommissioning plan.

1.7 Local Containmant.- Use of an enclosure and procedures to prevent or minimize the spread of contamination, often with IIEPA. filtered ventilation.

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

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

1.10 Reactor ShutdownFacility - Reactor shutdownfacility means that-there-is-no-fueHnthe containment structure located at the coreNeely Nuclear Research Center and everything contained inside the structure to include the reactor, support systems and remaining equipment.

1-H- ReactorGeinsoncnideactorromponentshalkneantmyspparatus-dcvicc or materiahhat-is1r-normal part-of-the- eaetorstructure--

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 restricted-areareactor facility in excess of the limits specified in 10 CFR Part 20.

c. An observed 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 decommissioning-activities performed under the USNRC possession only license..

1.13 Surveillance Freauency - Unless otherwise stated in these specifications, periodic Page 2 of 17-

.__ _ y_..

.- . _ . - ~-- . - . -. .

L surveillance tests, checks, calibrations and examinations shall be performed within the specified surveillance intervals. Tht e intervals may be adjusted plus or minus 25%. In cases _where the elapsed interval has exceeded _100% of the specified interval, the next i surveillance interval snali commence at the end of the original specified interval.

1.14 Surveillance Interval - The surveillance interval is the calendcr time between wrveillance tests, checks, calibrations and examinations to be performed upon nny instrument or component when it is required to be operable. These_ tests _ may be waived when the instrament, component or system is not required to be operable, but the instrument, component or system shall be tested prior to being declared operable.

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 shutdowa at all times and that the safety limits will not be exceeded.

SPECIFICATIOE All nuclear fuel remains removed from the reactor and contaimnent building. No fuel shall be reintroduced into these areas.

BASIS The reactor can not be operated without nuclear fuel.

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

OBJECTIVE t

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4 To specify the lowest acceptable level of performance or the minimum number of acceptable components for the safety related instrument & tion and systems as determined by the Technical Safety Review Committee.

SPECIFICATION The reactor shall not be made critical. The safety related incrumentation and system: sha11 be monitored to ensure that radiological releases to the environment are within acceptable levels as specified in 10 CFR Part 20.

DASIS Radiationmonitoring-will-assurethatereas-throughout-the-faeility4nsvhich4tigh radir. don arcastovidweur due-to-improper 4arnple-handlingrequipment-or-shielding 1novements are-identified: Table 2.1 provides a list of safety related instrumentation that will be maintained during the licensed period.

TABLE 2.1 SAFETY RELATED INSTRUMENTATION REQUIRED FOR LICENSED ACTIVITIES Instrumentation Set Point Mininmm Function Number Required Building Radiation < 10 mR/hr.or 3 (a) Alarm Monitor 2x background Tritium Monitor (c) Based on annual 1(b) Alarm & initiate containment calibration isolation Moving filter Based on annual 1(b) Alarm & initiate containment particulate calibration isolation Filter bank monitor Based on annual 1(b) Alarm & initiate containment calibration isolation (a) Area monitors shall be located on the experimental level, the reactor top and in the reactor facility basement.

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

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

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BASIS l Radiation monitoring will' assure that areas _throughout the reactor facility in which high ,

radiation areas could occur;if conditions; change-during Llicensed activities,1such as i improper sample handibg, equipment removal or shielding movements, are identified._

}

The fiker ban!c monitor, tritium monitor and the moying air particulate moniter provide diverse and redundant dennels which monitor particulate and gaseous releases from the  ;

reactor facility.  !

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2.3 - CONTAINMENT BUILDING APPLICABIL.lTI  ;

This specification applies to the reactor facility containment building requirements as long as heavy weter (nat 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 react _or facility.

SPECIFICATION t

. Containment integrity shali_ be maintained ~during licens6d activities which could cause airborne contamination'!evels to exceed 10CFR20,1 Appendix B, Table 2, Column 1, "Emuent Concentrations";.

BASIS Building containment is.a major. engineered l safety feature which serves as.the-final

, physical barrier to contain radioactive particles and ases b present or_ produced during licensed activities. Local control,will be usedlwithin the building when necessary to provide an additional contamination control boundary, '

The gasemitorNUREG-1537, fikerbank monitorPart2, karechamber" Guidelines for -  ;

Preparing and :hc ineving air particulatc incniter previdc divcisc nra redundant channcis- -

which inaniter par;icula:c and gescsus iclcescs from-theReviewing Applications for Licensing of.Non-Power Reactors", identifies the main purpose of a containment is to mitigate accident. consequences for the operating reactor building, Page 5 of 17

t' 2d GONTAINMENT-IWIL-DING APPLICABILIT-Y ThismpadScation-sppliesioihe-Georgia-Tech-Research-Reactor-(GTRR)The reason for maintaining containment buildingtequirementsintegrity during the possession only licensed -

period is to maintain the ALARA concept for off-site releases.

OBJEGTIVE To-minimizeihe-releaseof-sirborne-radioactive-materials-from1heGTRR :-

SPEC-IFIGATION Containment-integrity-shall-be-maintained-tluring decontamination-activities-involving possible-airbornewntamination icvcis abovc background-BASIS Building,:ontainment-ts-aThe major engineered-safety-feature-whieh-scrvcs asihefm' al physical-barrict to contain-radioactive particics and gascs picscat or produced-viuring-decontamination $etivitiessource of 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. Local-containmentvill bc uscd whcrc possible-to provide-an-additional-contamination-controlkmndaryOnce the heavy. water is removed from the reactor facility, a major source of tritium is eliminated.

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

QBJJICTIVE 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 Page 6 of 17

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within 'the limits of 10 CFR Part 20. f SPECIFICATIONS .

a. Liauld Wastes ;

(1) The concentration ~of gross radioactivity, above background, discharged

'from the Rcec;cs Buildingreactor; facility to.the sanitary sewer during

dccenterniae:ionlicensed 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 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 meet criteria for release following -

A processing, as an option, the liquid can be treated and solidified for future disposal at a licemed waste disposal facility.

(b. Gaseous Effluents; A'

(1) ' The maximum release rates of gross radioactivity in gaseous effluents -

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

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e (2) During decontaminationlicensed 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.

c. I.iquid Ef0uents (1) The maximum release rates of gross radioactivity in liquid ef0uents during decommissioninglicensed activities shall not exceed the limits set forth in 10 CFR Part 20, Appendix 3, Table 2, Column 2 and Notes.

(2) During decommissioninglicensed activities involving possible procluction 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.

DASIS

a. IJquid Wastu The liquid waste handling syrtem is described in the Safety Analysis Report dated January 10, 1995. Radioactive-effluent s-released-to-t he-sewage-on-the- basis-of gross-radioactivity-arc assumed-not-to-contain-Iodine-129-and-radium:-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 efnuent 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. a seous Effluents C

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 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 22.5a sector having the least diffusion on an annual average.

Isolation of the exhaust efnitent stack is initiated by high radiation in the off-gas Page 8 of 17

system.- Such isolation is required for potentially abnormally high gross radioactivity releases due to dcccan issioninglicensed activities while heavy water (not to include residual holdup)_is stored lat the reactor facility, 3.0 SURVEILLANCE REQUIREMENTS 3.1 SAFETY RELATED INSTRUMENTATION APPLICABILITY 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.

t

! BASIS l

Galibration-of-the-safety telated-instrumentatic i racans to mcasurc thc aciferinancc as gnidcd-bythe-vendorHnstructionstad pcifcimancc spccifications of thc4nstrumem-in-its icspensc-te accurately picscribcd-input signals.

TABLE 3.1 SURVEILLANCE REQUIREMENTS FOR SAFETY RELATED INSTRUMENTATION Surveillance Requirements Page 9 of 17 1

Channel Daily Source Check Known Parameter Check-(t) Semiannually Source Calibration Annually Kannetxhaust-gasBuilding x (1) - x GMigas-monitorx Radiation Monitor Tritium monitor x (2) x x xMoving filter particulate x (2) x x 41)-Applicable-onlyhm x (2) .x x decommissioning-activities are-beingterformed: Filter bank monitor-(1)_ Applicable only:when licensed activities are being performed which could cause a high radiation area.

(2). Applicable.only when licensed activities are being performed which could c "se an airborne release to the containment.

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 prescribed input signals.

3.2 CONTAINMENT BUILDING l

APPLICABILITY This specification applies to the surveillance of the containment building as long as heavy water (not including residual hold ip) 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.

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SPECIFICATION i

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 buildina, shall _be performed annually _at a pressure of at least 2.0 psig. Leakage 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 building 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 2n" Guidelines for Preparing and -

Reviewing Applications;foriticensinglof Non-PoweriReactors",; identifies the; main purpose of,a containment _is to mitigate accident coa *Tm for the operating reactor.

The reason for maintaining containmentjintegrity 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 oi 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.

Containment building isolation is initiated by a signal from either the tritiunt monitor, the 7

moving particulate monitor r a manual' push button on _the reactor co.nsole. Operability and trip point checks semi-annually will be performed =in,accordance with Table 3.1.

The containment has been leaked tested annually since 1963 and the leak rate of 0.5 % of the ouilding 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 exceedcd(No trend has dev_ eloped which j 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 build.ing or its penetrations will be tested to verify that such work has not adversely affected the leak tightness of the building.

t-4.0 SITE DESCRIPTION SPECIFICATf0N l a. The reactor facility is located on the Georgia Institute of Technology campus in the l city of Atlanta, Georgia.

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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 GTRRreactor facility is established as radius of 523 meters (1750 ft) from the reactor center.

5.0 ADMINISTRATIVE CONTROLS 5.1 ORGANIZATION

a. The organization for the management and oversight of-decommissioning operationsfor the reactor facility during the time when the possession only license is in effect shall be asas indicated in Figure 5.1. The Director, Neely Nuclear Research Center (NNRC) shall have overall responsibility for direction and operationperformance 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 Figure 5.1 Page 12 of 17

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b. The Radiation Safety Officer shall advhe the Director, Neely Nuclear Research Center in matters pertaining to radiological safety. lic/she has access to the Vice-Provost, Research and Graduate Studies and/or the President of the Institute as President Georgia Institute of Technology Dean Technical Safety Vice Provost for Research College of Engineenng Review C<>mmittec & Dean of Graduate Studies

.. .. Radiation Safety Neely Nucl at t arch Center needed.

5,2 TECIINICAL SAFETY REVIEW COMMI'ITEE

a. The Technical Safety Review Committee shall be estabbsheJ by the President of the Institute and shall be responsible for mentaintig health and safety standards associated with the decommissioning-oflicensed activities during the possession only licensed period for the reactor and-tissociated-facili@sfacility.

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 on as circumstances warrant. Written records of the proceedings, including any recon mendations 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 Page 13 of 17

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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 decommissioninglicensed activities. Minor modi 0 cations 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 modincations at its next scheduled meeting.

(4) Review and approve proposed changes to the Technical Specifications and license excluding organizational structure. The responsibility an 3 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 Speci0 cations, license or internal procedures or instruc. ions having safety signincance.

(7) Review decommissioning-cMurveillanceabnormalities occurring during licensed activities having safety signincance (8) Review audit reports.

(9) Audit decommissioning-and-surveillancerecords maintained during the licensed period for compliance with internal rules, procedures, and regulations and with licensed provisions including Technical Speci0 cations 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 in the :eactor facility equipment, systems, structures or methods of operatina that affect safety, at least once per calendar year (interval betvven audits not to exceed 15 months).

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e 5.3 PROCEDURES

a. The facility 1 hall telecommissioned1mdLicensed activities shall bc performed in accordance with approved written procedures. All decommissioningprocedures for licensed activities and major changes . ereto shall be reviewed and approved by the Technical Safety Review Committee prior to being effective. Changes which do not citer the original intent of a procedure may be approved 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 sy tems 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) Fhysical security of the facility.

(7) Preventive and corrective maintenance of systems related to safety.

5.4 OPERATING RECORDS

a. in addition to the requirements of applicable regulatiens, and in no v/ay substitutag therefore, records of the following items shall be maintained far a period of at least five years or as long as the facility is licensed by tile U.S.

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

(1) Normal facility operation and maintenance.

(2) Reportable occurrences.

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

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

(5) Off site environmental monitoring surveys.

(6) Facility radiation and contamination 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 experure for all personnel shall be kept indefinitely or until i the USNRC authorizes Pelt disposal.

5.5 ACTION TO HE TAKEN IN TIIE EVENT OF A REPORTABLE OCCURRENCE i

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

a. Origoing 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.

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.

546- REPORTINCrREQUIREMENTS In-addition-to-the-requirements-of-applicable-regulationsr-antHn-no-way-substituting-thereforerreports1 hall 4xMnade-toihe-USNRC-as-follows-5.6 REPORTING REQUIREMENTS in addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made to the USNRC as follows:

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a. Reportable Occurrence Report 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 and fax a the DirectorNRC's Operation Center Office-of-Lelear-Reactor-)<egulation-with-n copy-to the-Office-of-the-Regional-AdministratorrRegion-lito be followed by a written report within 1014 days to the Director;-Office-of-Nuclear-Reactor Regulation with nopy-to the Officeof-the Regional-AdministratorrRegion llU.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 and fax notification shall:

(1) describe, analyze and evaluate safety implicotions; (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. IJntinu!LEvents 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, i

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End of Document l

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