ML19340E962

Tech Specs 1.0-7.0 for Nuclear Reactor Facility License R-78

ML19340E962
Person / Time
Site:	05000148
Issue date:	03/31/1980
From:	KANSAS, UNIV. OF, LAWRENCE, KS
To:
Shared Package
ML19340E951	List: ML19340E959 ML19340E962 ML19340E968 ML19340E969
References
NUDOCS 8101160398
Download: ML19340E962 (31)

Text

_.

C[b h l

l Technical Specifications  ;

for the Nuclear Reactor .

Facility License R-78 .

The University of Kansas Departnent of Chemical and Petroleum Engineering Lawrence, Kansas March, 1980 G

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y.101160 M E aar s

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'== 1.0 DEFINITIONS T5 terms Safety Limit (SL), Limiting Safety System Setting (LSSS),

and Limiting Condition of Operation (LCO) are as defined in 50.36 of 10 CFR Part 50.

1.1 Safety Channel - A Safety Channel is a measuring or protective channel in the reactor safety system.

1.2 Reactor Safety System - The Reactor Safety System is a combination

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l of safety channels and associated circuitry which forms the auto-matic protective system for the reactor, or provides infomation

! which requires the iriitiation of manual protective action.

t 1.3 Ooerable - Operable means a component or system is capable of performing its intended function in its normal manner as demonstrated l

3 by continued nomal functioning in routine reactor operation and

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satisfactory completion of checks, tests, calibrations according to l

the specified schedule.

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

1.5 Channel Check _ - A Channel Check is a qualitative verification of acceptable performance by observation of channel behavior. This l

verification may include comparison of the channel with other independ-l ent channels or. methods m?asuring the same variable.

1.6 Channel Test - A Channel Test is the introduction of a signal into i the channel to verify that it is operable. - -

1.7 . Channel Calibration - A Channel Calibration is an adjustment of the i channel components such that its output responds, within acceptable Eh

= range and accuracy, to known values of the parameter which the channel M Chao

'2 bN measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip.

1.8 Unscheduled Shutdown - An Unscheduled Shutdown is any unplanned shutdown of the reactor, after startup has been initiated.

1.9 Reactor Shutdown - The reactor is shut down when the negative reactivity of the cold, clean core, including the reactivity worths of all ex-periments is equal to or greater than the shutdown margin. ,

1.10 Reactor Ooerating - The reactor is considered to be operating whenever it is not secured nor shutdown.

1.11 Reactor Secured - The reactor is secured when:

a. The core contains insufficient fuel to attain criticality under optimum conditions of moderation and reflection, or The moderator has been removed, or 5-b.

c. (1) Minimum number of control rods fully inserted as recuired by Technical Speicifcations, and (2) The withdraw buss key switch is in the off position and the key is removed from the lock, and (3) No work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods, and l (4) No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller.

1.12 True Value - The True Value of a parameter is its actual value at any instant. y

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3 Ib 1.13 Measured Value - The Measured Value of a parameter is as it appears on the output of a measuring channel.

1.14 Measuring Channel - A Measuring Channel is the combination of sensor, lines, amplific-s, and output devices which are conrcxted for the purpose of measuring the value of a parameter.

1.15 Reportable Occurrence - A Reportable Occurrence is any of those conditions described in Sect.on 6.5.3 of this specification..

1.16 An Exoeriment - An Experiment is an apparatus, device or material, placed in the reactor core, in an experiment facility, or in line with a beam of radiation emanating from the reactor, excluding devices designed to measure reactor characteristics such as detectors .and foils.

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'"" Secured Experiment - Any experiment, experiment facility, or a.

component of an experiment is deemed.to be secured, or in a ,

secured positio . if it is held in a stationary position relative to the reactor core. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, or other forces which are normal to the operating environment of the experiment (or, by forces which can arise as a result of credible malfunctions).

b. Movable Experiment - A movable experiment is _ne which may be inserted, removed, or manipulated while the reactor is critical. .

c. Untried Experiment - is a single experiment or class of ex-

'2F periments that has not been previously evaluated and approved by the Nuclear Reactor Committee.

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, dei l 1.17 C 1

i Experiment Facilities - An Experiment Facility is any structure, I

device or pipe system which is intended to guide, orient, post-tion, manipulate, control the environment or otherwise facilitate a multiplicity of experiments of similar charac'ter.

1.18 Control Rod - A control red is a rod fabricated from neutron absorb-ing material which is used to compensate for fuel burnup, tempera-ture, and poison effects. A control rod is magnetically coupled to its drive unit allowing it to perfom the safety function when the magnet is de-energized.

1.19 i

1 Readily Available on Call - Readily available on call means an individual i

who, (1) has been specifically designated and the designation known to the operator on duty, (2) keeps .'he operator on duty informed of where he may be rapidly contacted (e.g. by phone, etc.) (3) is capable of r@-

1 s getting to the reactor facility within a reasonable time under normal conditions.

1.20 Scram Time - is the elapsed time between the instant a limiting safety 6

system set point is reached and the inst. ant that the slowest control rod is fully inserted.

l 1.21 Safety Limits - are limits on important process variables which are found to be necessary to reasonably protect the .ntegrity of certain l

physical barriers which guard against release of radioactivity. The principal physical barrier is the fuel cladding.

2.0 Safety Limits and Limiting Safety System Setting 2.1 Safety limits of reactor coeration l

2.1.1 Limits of free convection

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Acclicability - This specification applies to the themal variables

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• affecting the core.

Power in Kilowatts Objective - To assure fuel cladding integrity and limit inventory of fission products.

Specifications -

The maximum reactor power shall be 250 Kw but the average power level will be limited to 10 Kw. Any one run at power levels above 10 Kw will be limited to an integrated energy of 750 Kw hrs followed by a recovery per!cd of such duration that the power averaged from the beginning of the run, including shutdown periods will be less than 10 Kw. The reactor may be operated at power levels up to 10 Kw during the recovery period.

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. Bases

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At a power level of 250 Kw the maximum temperature of the fuel plates is well below the safe cladding temperature of MTR elements.

2.2 Limitina Safety System Setting 2.2.1 Safety Channel Set-points Applicability - This specification applies to the set-points of the safety channels.

Obiective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded.

Specification - The safety channels shall be set to scram at 150% of full power or less and the Linear Channel shall be set to initiate shim rod reversal at 120% full power or less. ,

Bases - The set points are chosen to avoid boiling in the core during s routine operation, and to assure the cladding temperature will be well

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below the melting point of the cladding.

bI-3.0 Limiting Conditions for Ooeration l 3.1 Reactivity Limitations i

I 3.1.1 Shutdown Margin The minimum shutdown margin provided by the control rods in a Cold xenon-free condition with the maximum worth rod fully withdrawn and with the highest worth non-secured experiment in the core in its most reactive state shall not be less than 0.5%AK. This specification assures the reactor can be shutdown even if its most reactive rod is stuck in its fully withdrawn position.

3.1.2 Excess Reactivity The core shall not be loaded with an excess reactivity greater than 1.5%.

3.1.3 Exceriments Reactivity limits on experiments are specified in 3.5 below.

3.2 Control and Safety Systems 3.2.1 Scram Time The scram time after receipt of a scram signal shall not exceed one second.

3.2.2 Measuring Channels The minimum number and type of measuring channels operable and providing information to the reactor operator required for operation are given as follows: .

Linear Power channel Log N channel

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• Log count rate channel vs l

Rod position indicators

• required during startup.

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Bases - The count rate channel covers the neutron flux range from 5

the source level of 2 cps to 10 cps on a logarithmic scale. It enables the operator to start the reactor safely from a shutdown condition, and to bring the power to a level that can be measured by the Log N channel.

The Log N and the Linear Power channel provide redundant information on reactor power from .25 W to 250 Kw.

Rod position indicators show the operator the relative 'positions of control rods and enable rod reactivity estimates to be made.

3.2.3 Safety Channels The minimum number and type of channels providing automatic action that are required for reactor operation are as follows:

g Channel Function Safety Scram, P7375 Kw

Log Count Rate Inhibit, CR42 Cps Inhibit, Per<20 see Reverse, Per<10 sec Log N Reverse, Per<10 see Scram, P>500 Kw Inhibit, Per420 sec Scram, Perf2 see Linear Scram, P>375 Kw ,

Reverse, P>300 Kw Manual Scram Scram Key Switch Inhibit ,

Bases - The power level scram provides redundant automatic protective l action to prevent exceeding the safety limit on reactor power. 4 1

The period scram, assisted by the period reverse and rod inhibit, l limits the rate of increase in reactor power to values that are I i

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controllable without reaching excessive power levels or temperatures.

These functions are not limiting safety system settings.

The inhibit on the count rate channel prevents inadvertent criticality during cold startup that could arise from lack of neutron information or from too rapid reactivity insertion by control rods.

The over power reverse provides automatic action to reduce power and minimize the chance of incipient boiling in the core.

l Bypass is pernitted on those parameters that can be monitored by alternate means if the initiating circuit malfunctions.

The key switch prevents unauthorized operation of the reactor.

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3.3 Radiation Monitoring System f (1) During reactor operation and handling of highly radioactive materials, the reactor bay shall be monitored by at least two of the four area monit7rs whose locations are listed below;

a. East wall

b. South wall

c. West wall near demineralizer l

d. Above the reactor tank

' For the reactor to be operated above 10 kW, only the monitor on l

either the east or south u ll may be out of service. The monitors shall provide a readout and a signal which activates the annunciator and ' Radiation Level' warning light on the conso;e.

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In the event that fewer than two of the four area monitors are in service, reactor operation at power levels no greater than 10 kW may continue using portable gamma sensitive monitoring devices with the following restric-tions:

a. The monitoring devices used must be verified to be operable just prior to first use of the day using a check source.

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b. At least two of the areas listed in a. to d.

above shall be monitored by the installed and/or the portable monitoring equipment. The portable gamma sensitive detectors shall be monitored d!@ continuously during reactor operation and during sample removal.

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c. Radiation levels and locations shall be recorded in the log book.

(2) An air monitor shall be placed in the reactor bay at frequent time intervals to determine the amount of radioactivity present in the air.

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i i d!$l The reactor bay area shall be continuously monitored if the evaluation af any experiment shows that 25% of the allowable I

exposure as described in Table I, Appendix B of 10 CFR 20 can be exceeded under accident conditions.

(3) The power level of the reactor shall be limited to a value such that the radiation level at the foot of the drive at the south end of the building will be less than 2 mrem /hr. Whenever a radiation level above 2 mrem /hr exists on the drive, the drive I

will be posted as a radiation zone to limit free access to the drive. -

3.4.3 Containmen_t

The containment ventilation system is set up as follows:  !

Air is circulated by a ventilator on the ceiling. When the ventilation system is shutdown, outside vents automatically close. It is possible to isolate the reactor bay from the remainder of the building by the closing of doors.

3.5 Limitations on Experiments

a. Applicability - This specification applies to those experiments in the reactor and its experiment facilities.

b. Objective - The objective is to prevent damage to the reactor or excessive release of radioactive material in the event i

of an experiment failure and also to prevent the safety limits from being exceeded. ,

c. Soecification - Experiments installed in the reactor shall meet the following conditions:

(1) The reactivity worth of any individual experiment (u[=

shall not exceed 1.5% delta k/k.

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11 Ehl (2) The sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities at the time shall not exceed 1.5% delta k/k. This includes the total potential reactivity insertion which might result from experiment malfunction, accidental

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experiment flooding or voiding, accidental removal or insertion of experiments.

(3) If estimated worth of experiment is greater than 1% delta

. k/k, the actual worth shall be measured and recorded at the time'of initiial insertion of'the experiment.

(4) Failure of an experiment shall not lead to a direct

. failure of a fuel element or other experiments.

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(5) Where the possibility exists that the failure of an i:2 experiment could release radioactive gases or aerosels to the reactor bay or' atmosphere, the quantity and type of material shall be limited such that the airborne concentration of radioactivity averaged over a year will not exceed the limits ofTabkeIIofAppendixBof10FR20 assuming 100%ofthegases or aerescis escape.

(6) Liquids contained only in breakable containers shall not

. be irradiated. .

(7) All samples of soluble material being irradiated within the pool shall be encapsulated in a water tight non-breakable radiation resistant container.

(8) All material being . irradiated within the thermal column or beamports which is capable of contaminating local areas will E5h be placed in dust proof containers.

(9) Known explosive materials such as gunpowder, dynamite,

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TNT, or nitroglycerine shall not be irradiated in the reactor or experimental facih ties.

d. Bases (1) The limit on reactivity limits the severity of power excursion. S- :iSR page 36 (Ref 1) 3.6 Fuel

a. Apolicability - These specificaticas apply to the number and condition of the fuel elements in the core.

b. 0bjective - To ensure that power is distributed in the core among a sufficient t number of fuel elements to avoid excessive peak / average ratio, and to avoid excessive release of fission products.

c. Soecifications (1) The number of fuel elements loaded will be such as to assure the Ky is 1.5% or less.

(2) Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. An abnormal increase in radiation level at the demineralizer together with detection of fission products in tne pool water shall constitute initial evidence of cladding rupture and require identification of the cause.

(3) The fuel elements shall be MTR type.

d. Bases (1) This specification limits the number of fuel elements that can be loaded into the core.

(2) There is a normal small and variable amount of fission _

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product release due to uranium contamination in the coolant 1

E3 lll[; and on fuel plates. It is thus safe to specify a reccgnizable and substantial increase in background as possible of indica-J tion of cladding rupture.

(3) This specification defines the type of fuel used in the reactor.

3.7 pool water quality

a. Acolicability - This specification applies to the reactor pool l .

water in contact with the fuel elements. -

b. Objective - To minimize corrosion of the fuel element cladding, and

'o t prevent the, activation of dissolved materials.

c. Saecification (1) The recirculated pool water shall be tested for resistivity before the first reactor start up of the first reactor start-5bI ing of the day. Assistivity shall not be below 1 megchm centimeter averaged over a month.

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l (2) .e reactor shall not be operated when water temperature i

C i measured at the surface exceeds 120 F.

! d. Bases

- Water purity is controlled to creserve the integrity of the fuel element. The temperature is limited to prevent boiling.

4.0 Surveillance Requirements 4.1 General The requirements listed below generally prescribe tests or inspections to verify periodical.ly that the performance of required systems is in accordance with specifications given above. In all instances where j

the specified frequency is annual, the interval between tests is not

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• = to exceed 14 months; when semiannual, the interval should not exceed

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7 months; when monthly the interval shall not exceed 6 weeks; when weekly the interval shall not exceed 10 days; and when daily the interval shall not exceed 3 days.

4.2 Safety Channel Calibration A channel calibration of each safety channel shall be performed annually.

I 4.3 Reactivity Surveillance (1) The reactivity worth of each control rod (including the regulating rod) and the shut-down margin shall be determined when-

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ever operation requires a re-evaluation of core physics parameters.

The rod worth will be detennined using the reactivity-period or

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l red-drop methods.

(2) The reactivity worth of an experiment shall be estimated, or P

! measured at lw power, before conducting the experiment. e (3) The control rods shall be inspected every six years or whenever drop times exceeding .9 seconds are observed.

4.4 Control and Safety System Surveillance (1) The scram time shall be measured annually. If a control rod is removed from the core temporarily, or if a new rod is installed, its drop time shall be measured before reactor operation.

(2) A channel check of the power level measuring channels shall be performed daily whenever the reactor is in operation. A channel check before startup is, however, required on any channel receiving

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maintenance during the shut-down period.

1 4.5 Radiation Monitorino System The area monitors shall be calibrated annually The area monitors shall receive a channel check and a setpoint verification daily during reactor operating periods.

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15 61 5 4.6 Reactor Fuel Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifications.

Each new fuel element will be inspected for damage and flow obstruc-tiens prior to insertion into the core.

4.7 Sealed Sources The PuSe sealed source shall be leak tested semiannually.

5.0 Design Features Those design features relevant to operation safety and to limits thai:

have been previously specified are described below. These features shall not be changed without. appropriate review.

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=== 5.1 Reactor Fuel Fuel elements shall be of the general MTR type with thin plates 235 U and r, lad with aluminum.

containing fuel enriched to about 93%

Elements shall confonn to these nominal specifications:

Overall size 3 in x 3 in x 35 in Clad thickness .020 in Plate thickness .060 in No. of Fuel plates 10 standard element 5 control rod element Plate Attachment r pinned Fuel content Swagedg3,U/fullelement 167 g

• I 5.2 Control and Safety Systems Design features of the components of the system 3.2.2, 3.2.3 that _

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are important to safety are 'given below.

5.2.1 Power Level Safety Channels

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== For this functica two independent measuring channels are provided.

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(Th Each channel covers reliably the range from about .25W to 250Kw.

Each channel comprises a compensated boron-coated ion chamber feeding an amplifier that controls electronic switches in the DC current that flows through each control rod electromagnet.

Each channel controls and scrams all shim rods. Each channel is fail-safe. Each channel indicates power level on a panel meter 1

I allowing channel checks to be done during reactor operation. Each chamber can be changed in position, over a limited range, so as to i allow the channel reading to be standardized against reactor thermal l power. One channel has a logarithmic output indication on both a panel meter and a chart recorder. Rate of change of power information is  !

also derived, in the form of a period, that can produce a fast scram  !

or rod reverse. gg.

W 5.2.2 Count Rate Channel

- A fission chamber is used to supply pulses to an amplifier and logarithmic count rate circuitry. pulse height discrimination selects pulse amplitudes that correspond to fission events and rejects those from alpha particles. Count rate on a logarithmic scale is displayed on a panel meter and a chart recorder. The channel covers a reege of 1-105cps. The fission chamber is raised when the counting rate 5

approaches or exceeds 10 cps. ~To prevent control rod withdrawal when i

the neutron count rate information may not be reliably indicated, inhibits are provided on count rate. A scaler is also provided for obtaining accurate values at low count rates if needed (e.g., approach to critical with new fuel or new core configuration).

5.2.3 Neutron Source Fcr obtaining the reliable neutron information necessary for startup

17 ih s

from a cold shut-down condition, a. Plutonium-Beryllimum neutron 4

source is provided for insertion into the core as needed. Integrity l

of the source is checked by periodic wipe tests.

5.3 Rod Control System 5.3.1 Control Rods Three control rods are provided for the control of core reactivity.

These rods contain boron-carbide. Individual integral worth.s vary from about 1-4% AK, depending on position and core configuration.

The rods are coupled to drive shafts through electromagnets that allow quick release of the rods. Position indicators on the control console show the extent of withdrawal for each rod. To limit the

' rate of reactivity increase upon startup, the rod drive speeds are 2_:~;

limited to 5 in/ min.

5.3.2 Regulating Rod One control rod serves as the regulating rod for fine control and maintenance of constant reactor power for long periods.

5.4 Cooling System .

5.4.1 Primary Cooling System Core cooling is provided by natural convection through the reactor core. Water flow is from the reactor pool, through the demineralizer, and back to the pool, at a flow rate of approximately 3 gal. per minute.

5.5 Containment System The reactor is housed in a building of reinforced concrete with an 2

exterior stone facing and a floor space of approx.15,000 ft . The q:. circulatory fan for the reactor bay has its shut off switches near

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I the control c.)nsole and is equipped with automatic closure devices.

5.6 Fuel Storage l 1

5.6.1 Fuel Storage and Transfer The storage pool is located below the floor level of the reactor bay and is capable of storing the complete fuel inventory. The geometry of the storage racks are such that criticality is impossible when the complete fuel inventory is properly stored there. '

6.0 Administrative Controls 6.1 Oreanization 6.1.1 Structure ,

The organization for the management and operation of the reactor facility shall be as a minimum the structure shown in Tig.1. Job -

titles shown are for illustration and may vary. Three levels of authority are provided, as follows:

Level 1: Individual responsible for site administration.

Level 2: Individual responsible for the reactor facility license, operation, management and daily reactor operations, Level 3: Reactor operating staff.

f The Nuclear Reactor Committee shall report to Level 1.

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i 6.1.2 Responsibility Responsibiility for the safe operation of the reactor facility shall be within the chain of comand shown in Figure-1. Management levels in addition to having responsibility for the policies and operation of the reactor facility.shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for h

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FIGURE 1 Level 1 Dean of Engineering .

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Level 2 Diredtor i

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Level 3 Operating Staff kh e

6 m.

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adhering to all requirements of the operating license and technical specifications. In all instances responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

6.1.3 Staffino

a. The minimum staffing when the reactor is not secured shall be:

l (1) A licensed Reactor Operator in the control- room.

(2) A second person present at the reactor facility.

(3) A licensed Senior Reactor Operator shall be readily available on call.

(4) A member of the operating shift shall be designated by Level 2 management as knowledgeable in radiation control.

b. Events requiring the presence of a Senior Operator:

(1) All fuel-element or control-rod alternations within the reactor core region.

(2) Relocations of any experiments with reactivity worth greater than 0.4%.

'(3) Recovery from unplanned or unscheduled shutdowns unless they are of a type excluded by the Level 2 authority.

Such exclusions shall be posted in the control room or placed in l the appropriate procedures. Furthermore, the presence of a senior operator at the facility shall not be required during recovery from unplanned or unscheduled shutdown or significant reduction in power in instances which result from:

1. Electrical power interruptions from internal or external failures exclusive of power supply failures of the reactor b instrumentation control and safety systems;

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2. False signals, which, in the opinion of the Senior Operator, were properly verified to be false and to have resulted from monitoring, experimental, or control equipment, or from personnel inadvertence; and a

3. Intentional shutdowns made by the Reactor Operator which are not related to the safety of the reactor; provided that prior to the initiation of such recovery, the Senior Operator shall be notified of the shutdown of power reduction, and shall determine that the shutdown was caused by one of the enumerated occurrences, and snali determine that his presence at the facility during recovery is not required.

6.1.4 Selection and Trainina of Personnel

.a The selection, training, and requalification of personnel shall meet or

Q exceed the requirements of Appendix A of CFR Part 55 and be in accordance with the requalification plan approved by the Commission.

6.1.5 Review and Audit The independent review and audit of reactor facility operations shall be performed by a qualified person designated by the Nuclear Reactor Comittee.

6.1.5.1 Comcosition and Qualifications The Nuclear Reactor Comittee shall be composed of a minimum of 5 members. The members shall collectively provide a broad spectrum of expertise in science and engineering. Members and alternates shall be appointed by and report to the Level 1 authority. They may include individuals.from within and/or outside the operating organiza-(;II: tion. Qualified and approved alternates may serve in the absence of regular members.

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D.

6.1.5.2 Charter and Rules The committee shall function under the following operating rules:

a. Meetings shall be held not less than semi-annually or more frequently as circumstances warrant consistent with effective conitoring of facility activities,

b. A quorum shall consist of not less than one half the membership, where the operating staff does not constitute a majority.

c. Sub-groups may be appointed to review specific items.

d. Minutes shall be kept, and shall be disseminated to members and to the Level 1 authority within one month after the meeting.

The Committee shall appoint one or more qualified individuals to perform the Audit Function.

.62 95" 6.1.5.3 Review Function The following items shall be reviewed by the review group or a subgroup thereof:

a. Determinations that proposed changes in equipment, syste-s, tests, experiments, or procedures do not involve an unreviewed safety question.

b. All new procedures and major revisions there to having safety significance, proposed changes in reactor facility equipment, or systems having safety significance.

c. Tests and experiments in accordance with section 6.3.

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d. Proposed changes in technical specifications, license, or charter. j I

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

Violations of internal procedures or instructions having safety significance. kb: l

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15. f. Operating abnonnalities having safety significance, and audit-reports.

g. Reportable occurrences listed in section 6.5.3.

6.'t.5.4 Audit Function The audit function shall include selective but comprehensive examination of operating records, logs, and other documents. Where necessary, discussions with responsible personnel shall take place. In no case shall the individual or individuals conducting the ludit be immediately responsible for the area being audited. The following items shall be audited:

I a. The conformance of facility operations to the technical specifica-

. tions and applilable license or charter conditions, at least once per calendar year (interval not to exceed 18 months).

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• b. The retraining and requalification for the operating staff, at least once every other calendar year (interval not to exceed 30 months).

c. The results of actions taken to correct deficiencies occurring in reactor facility equipment, systems, structures, or methods of operations that affect reactor safety, at least once per year (interval not to exceed 18 months).

d. The reactor facility Security Plan and implementing procedures at least once every other calendar year (interval not.to exceed 30 months).

Deficiencies uncovered that affect reactor safety shall imediately be reported to the Level 2 authority. A written report of the findings of the audit shall be submitted to the Level 2 authority and the Nucibar~ Reactor Committee members within 90 days after

!5h the audit has been completed.

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24 6.2 Procedures E5

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There shall be written procedures for, and prior to, initiating any of the activities listed in this section. The procedures shall be reviewed by the Nuclear Reactor Committee and approved by Level 2 or designated alternates, and such reviews and approvals shall be documented. Several of the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures.

a. Startup, operation, and shutdown of the reactor.

b.

Fuel loading, unloading, and movement within the reactor.

c.

Routine maintenance of major components of systems that could have an effect on reactor safety.

d.

Surveillance tests and calibrations required by the technical specifications or those that may have an effect on reactor safety.

e.

Personnel radiation protection, consistent with applicable regulatio%57 ns.

f.

Administrative controls for operations and maintenance and for the conduct of irradiations ano experiments that could affect reactor safety or core reactivity.

g. Implementation of the Security Plan.

Substantive changes to the above procedures shall be made only after documented review by the Nuclear Reactor Committee and approval by Level 2 or designated alternates. Minor modifica-tions to the original procedures which do not change their original intent may be made by the Level 2 authority. Temporary changes to the procedures that do not affect reactor safety may be made I

by a Senior Reactor Operator and are valid for a period of one month.

Such temporary changes shall be documented and reported ,=..

to Level 2 or designaced alternate.  %'

. 25 h 6.3 Exoeriment Review and Approval

a. All new experiments or classes of experiments that could affect reactivity or result in release of radioactive materials shall be reviewed by the Nuclear Reactor Comittee. This review shall assure that compliance with the requirements of the license, technical specifications, and applicable regulations has been sat.sfied, and shall be documented.

b. Prior to review, an experiment plan or proposal shall. be prepared describing the experiment in:hding any safety considerations.

c. Review coments of the Nuclear Reactor Comittee setting forth any conditions and/or limitations shall be documented in Comittee minutes and submitted to Level 2 authority.

d. All new experiments or classes of experiments shall be approved 43 in writing by Level 2 or designated alternates prior to their initiation. ,

e. Substantive changes to approved experiments shall be made only after review by the Nuclear Reactor Comittee and written approval by Level 2 or designated alternates. Minor changes that do not significantly alter the experiment may be approved by the Level 2 authority.

f. Approved experiments shall be carried out in accordance with established approved procedures.

6.4 Reouired Actions 6.4.1 Action to be taken in Case of Safety Limit Violation ,

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a. The reactor shall be shutdown, and reactor operations shall not be resumed until authorized by the Comission.

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b. The safety limit violation shall promptly be reported to the Level 1 authority or designated alternates.

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26 .j s-

c. The safety limit violation shall be reported to the Comission in accordance with section 6.5.3.

d. A safety limit violation report shall be prepared. The report shall describe the following:

(1) Applicable circumstances leading to the violation.

(2) Effect of the violation upon reactor facility components, systems, or structures.

(3) Corrective action to be taken to prevent recurrence.

The report shall be reviewed by the Nuclear Reactor Comittee. A follow-up report describing extant activities shall be submittsd to the Comission when authorization is sought to resume operation of the reactor. ,

6.4.2 Action to be taken in the event of an occurrence as defined in section e pr 6.5.3, a-1, 3:

a. Corrective action shall be taken to return conditions to normal; otherwise, the reactor shall be shutdown and reactor operation shall not be resumed unless authorized by the Level 2 authority j

or designated alternates.

b. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates.

. c. All such occurrences where applicable shall be reported to the l Commission in accorcance with section 6.5.3.

d. All such occurrences including action taken to prevent or reduce the probability of a recurrence shall be reviewed by the Nuclear f

l Reactor Comittee.

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6.5 Reports In addition to the requirements of applicable regulations, reports shall be made to the Commission as follows:

6.5.2 Operating Reports Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted to the appropriate NRC Office within 3 months following the end of each prescribed year.

Each annual operating report shall include the following information:

a. A narrative summary of reactor operating experience including energy produced by the reactor.

b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence, but excluding those of

,g_ the types listed in Sect. 6.1.3.b(3) above.

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c. Tabulation of major preventive and corrective maintenance operations having safety significance.

d. Tabulation of major changes in the reactor facility procedures, and new tests and/or experiments significantly different from those performed previously and which are not described in the l

Hazards Summary Report, including conclusions that no unreviewed I

safety questions were involved.

e. The results of any environmental surveys performed.outside the facility.

f. A sunnary of significant (above 500 mrem) radiation exposures

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received by facil.ity personnel and visitors in any one year including the dates and times of significant exposures.

6.5.3 Special Reports (Reportable Occurrences),

1. ... :.

a. There shall be a report not later than the following working day ,

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by telephone and confirmed by telegraph or r,imilar cetayance= l

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to the Comission to be followed by a written report within 14 days of any of the following:

(1)

Release of radioactivity from the reactor above allowed limits.

(2) Violation of Safety Limits.

(3) Any of the following:

(a)

Operation with actual safety-system settings less con-servative than the limiting safety-system settings specified in the Technical Specifications.

(b) Operation in violation of Limiting Conditions for Operation established in the Technical Specificatio r.

(c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of 6f:

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performing its intended safety function unless the malfunction or condition is discovered during tests or periods of reactor shutdowns.

(Note:

Where components or systems are provided in addition I to those required by the Technical Specifications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)

(d) An unanticipated or uncontrolled change in reactivity greater than or equal to 1% AX/K.

I (e) Abnormal and significant degradation in reactor fuel.

I and/or cladding, coolant boundary, or containment boundary (excluding minor leaks) where applicable which could l I

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29 m:4=., ,

"" result in exceeding prescribed radiation-exposure limits of personnel and/or environment.

(f) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused an unsafe condition with regrad 4

to reactor operations.

b. A written report within 30 days to the Commission of:

(1) Permanent changes in the facility organization structure (2) Significant changes in the transient or accident analysis as described in the Hazards Summary Report.

6.6 Records 4

Records of the following activities shall be maintained and retained I

for the periods specified below. The records may be in the form of

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logs, data sheets, or other suitable forns. The required information may be contained in single, or multiple records, or a combination thereof.

Recorder charts showing operating parameters of the reactor for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum period of two years.

6.5.1 Records to be Retained for a Period of at least Five Years or for the ,

Life of the Component Involved whichever is smaller.

a. Normal reactor facility operations (including scheduled and unschedulo shutdowns). Note: Supporting documents such as checklists, log i sheets, etc. shall be maintained for a period of at least two years.

b. Principal maintenance operations.

c. Reportable occurrences.

mgg. d. Surveillance activities required by the Technical Specifications.

, eg

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e. Reactor facility radiation and contamination surveys where required by app 11cible regulations.

f. Experiments performed with the reactor.

g. Special Nuclear Materials (SNM) inventories, rer.eipts, and shipments.

h. Approved changes in operating procedures.

i. Records of meeting and audit reports of the Nuclear Reactdr Committee.

j. Sealed Source leak test results.

6.6.2 Records-to be retained for a,t least One Requalification Cycle or for the length of Employment of the Individual whichever is Smaller.

a. Retraining and requalification of licensed operations personnel.

However, records of the most recent complete cycl'e shall be maintained at all times the individual is employed. gr[

6.6.3 Records to be Retained for the Lifetime of the Reactor Facility: (Note:

Annual reports may be used where applicable as records in this section)

a. Gaseous and liquid radioactive effluents released to the environs,

b. Off-site environmental-monitoring surveys required by the Technical Specifications.

c. Radiation exposure for all personnel monitored.

d. Updated drawings of the reactor facility.

7.0 References

1. Hazards Summary Report (July,1959) g