ML20210R113

Amend 7 to License R-108,deleting Requirement for Scram Function on Min Reactor Period,Changing Periodic Insp of Reactor Fuel Schedule,Adding Defined Quarterly Surveillance Interval & Modifying Facility Organizational Structure

ML20210R113
Person / Time
Site:	Dow Chemical Company
Issue date:	08/19/1997
From:	Mendonca M NRC (Affiliation Not Assigned)
To:
Shared Package
ML20210R101	List: ML20210R097 ML20210R113 ML20210R122
References
R-108-A-007, R-108-A-7, NUDOCS 9709020312
Download: ML20210R113 (46)

Text

-= . _ _ - - _ . - - - - - -- _ - . . -. .

4 p##400q' f t UNITE] STATES NUCLEAR REGULATORY COMMIS810N f WASHINGTON D.C. 30MH001

"%...../

DOW CHEMICAL COM.PANY DOCKET NO. 50 264 AMENDMENT TO FACillTY OPERATING LICENSE Amendment No. 7 License No. R 108

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

A. The application for an amendment to facility Operating License No.

R-108 filed by the Dow Chemical Company (the licensee) on May 5, 1994, as supplemented on February 8 and August 7, 1995, June 18 and October 8, 1996, and February 24 and June 16, 1997, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended-(the Act), and the regulations of the Commission as set forth in Chspter I of Title 10 of the Code of Federal Reaula11oni (10 CFR);

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commissioni C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the regulations of the Commission D. The issuance of this amendment will not be inimical to the comT,w defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with the regulations of the Commission as set forth in 10 CFR Part 51, and all applicable requirements have been satisfied; and F. Prior notice of this amendment was not required by 10 CFR 2.105, and publit.ation of notice for this amendment is not reauired by ,

10 CFR 2.106.

[TP12!!MMpg c{' l V 1 "

_ ~ _

, v i

- 2-

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and amended paragraph to read2.C.(2)llows:of as fo Facility Operating License No. R-108 is hereby (2) Technical Soecifications l The Technical Specifications contained in Appendix A, as revised through Amendment No. 7, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical 1 Specifications.

3. This license amendment is effective as of the date of issuance. l FOR THE NUCLEAR REGULATORY COMMISSION

% \

ck  % j Marvin M. Mendonca, Acting Dir cto.

Non-Power Reactors and Decommissioning Project Directorate-Division of Reactor Program Management  ;

Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specification Changes  !

Date of issuance: August 19, 1997 I

s v a ,,,-,v - - + - , w ,w , $ w ,,, r v, - - -. , , , - , , - - , , - - + - - , - -- ,-,,,ms.,n,1,-r-nv---,:v>r,, - - - ,---+r,-' -, ,-,,,,w--, -.~~ * --,- - ~

4V i

ENCLOSURE TO LICENSE AM@QtlENT NO. 7 FACILITY OPERATING LICENSE NO. R-108 DOCKET NO. 50-264 Replace the following ) ages of the Appendix A Technical Specifications with the enclosed pages. T1e revised pages are identified by amendment number and contain vertical lines indicating tlie areas of change.

Remove Insert Cover Page Cover Page 1 43 1 42

., /'

i i

e 1

'ECHNICAL SPECIFICA110NS

{

i i FOR THE [

4 1 1

DOW TRIGA RESEARCH REACTOR 1

FACILITY LICENSE R.108 ,

i. .

AMENDMENT NO. 7 i

This document includes the Technical Specifications and the bases for the Technical Specincadons. The bases provide the r technical support for the individual Technical SpeelDeadons and j are included for informadon purposes only. The bases are not i

, part of the Technical Specificadoni and they do not consutute ,

!!mitadons or requirements to whleh the licernee must adhere. '

i  !

?

{ i O

F

"_=

4 s.

_ _ . .- e W L m-wy-, w,.w w . a. g-n-m w-els t .- y-. -- - y..- -, rr wr g y'w-^ntw r-v --

vf

o I, DEFINITIONS '

l.l. . A1. ARA . The ALARA (As Low As Reasonably Acidevable) program is a set of procedures which is intended to minimlic occupational exposures to ioniting radiation and releases of radioacuve materials to the environment.

1,2, Channel . A channel is a combinadon of sensors, electronic circuits, and output devices connected by the appropriate commtv.ications network in order to measure and display the value of a parameter, 1.3. Channel Calibration . A channel calibmdon is an adjustment of a channel such that its output wrresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip and shall include .i Channel Test.

1.4. Charmel Check . A chsanel check is a qualitadye verification of acceptable performance by observation of channel behavior. The verificadon shall include comparison of the channel with other independent channels or systems measuring the same variable, whenever possible.

1.5. Channel Test . A channel test is the introduction of a signal into a channel for verificadon of the operability of the channel, 1.6. Connnement Connnement is an enclosure of the facility which controls the movement of air into and out of the facility through a controlled patit 1.7. Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive posidon from the condition where the reactor is exactly critical, 1.8. Etneriment . An experiment is any device or material, not normally part of the reactor.-

which is introduced into the reactor for the purpose of exposure to radiadon,'or any operation which is designed to investigate non routine reactor characteristics.

'1,9. Exnerimental Facilities include the rotary specimen rack, sample containers replacing fuel elements or dummy fuel elements in the core, pneumatic transfer systems, the central thimble, and the area surrounding the core.

1.10. Facilltv Director . Person with line management responsibility to whom the Reactor Supervisor reports 'the person also serves as chairperson of the Reactor Operations Committee.

1.11. I imiting conditions for Oneration - Limiting Conditions for Operation (LCO) are i administratively established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor.

1.12. 1 imiting Safety System Settine (1 MSS) . An LSSS is the actuating level for automatic I protective devices related to those variables having signincant safety functions.

. l .13. Measured Value - A measured value is the value of a parameter as it appears on the l output of a channel.

Amendment Na 7

.i.

If '

1.14. Modified Routine Exneriments - Modilled roudne experiments are experiments which I have not been designated as roudne experiments or which have not been performed previously, but are similar to roudne experiments in that the hazards are neither significantly different from nor greater than the hazards of the corresponding roudne experiment.

1.15. Movable Exneriment A movable experiment is an experiment intended to be moved in j or near the core or into and out of the reactor while the reactor is operaung.

1.16. Onerable A component or system is operable if it is capable of performing its intended l Iunction.

1.17. Oneratine A component or system is operaung !f it is performing its intended funcdon. l 1.18. Radiation Safety Committee (RSC) 'the RSC is responsible for the administradon of all Dow Midland kication acuvides involving the use of radioactive materials and radiation sources including assuring compliance with US NRC reguladons.

1.19. Reactivitv I imits The reactivity limits are those limits imposed on reactor core excess I reactivity. Quantities are referenced to a Reference Core Condition.

1.20, Reactivity Worth of an Exneriment - The reactivity worth of an experiment is the I maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

I 1.21, Reactor Oneratine - The reactor is operadng whenever it is not secured or shutdown.

1 22. Reactor Safety Circuits - Reactor safety circuits are those circuits, including the l associated input circuits, which are designed to initiate a reactor scram.

1.23. Reactor Secured The reactor is secured whenever: l a) it contains hsuf0cient ilssile material present in the reactor, adjacent experiments or control rods, to attain cridcality under opumum available conditions of moderation and redection, or b) the console switch is in the "off" posidon, the key is removed from the switch, and the key is in the control of a licensed reactor operator or stored in a locked storage area; and sufficient control rods are inserted to assure that the reactor is subcritical by a margin greater than $1.00 cold, without xenon; and no work is in progress involving core fuel, core structure, installed control rods or control rod drives unless those drives are physically disconnected from the control rods; and no e<periments in or near the core are being moved or serviced that have, on movement. a reactivity worth exceeding $0.75.

Amendment No 7 2-

o 0

1.24. Renctor Shutdown De reactor is shutdown if it is subcritical by at least one dollar and l the reactivity worth of all experiments is accounted for, 1.25, Reactor Onerations Committee (ROC) - De ROC is charged with direct oversight of the l reactor operations, including both review and audit functions.

1.26. Renctor Safety Systems . Reactor Safety Systems are those systems, including l associated input channels, which are designed to initiate automade reactor protection or to provide information for inidation of manual protective acdon.

1.27. Reference Core Condition . De Reference Core Condidon is that condition when the core I is at ambient temperature (cold) and the reactivity worth of xenon in the fuel is negligible (less than $.30).

1.28. Resentch Renctor. A Research Reactor is a device designed to support a self sustaining l nuclear chain reacdon for research, development, education, training, or experimental purposes, and which may have provisions for the production of radioisotopes.

1.29. Renortable Occurrence . A Repartable Occurrence is any of the following which occurs l during reactor operation:

a) Operation with actual safety system settings for required systems less conservadve than the limiting safety system settings specified in Technical Specificadon 2.2.

M Operation in violation of limiting conditions for operation established in the Technical Specifications.

c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety funcdon unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.

d) Any unanticipated or uncontrolled change in reactivity greater than one dollar.

. Reactor trips resulting from a known cause are excluded.

1 Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary e) which could result in exceeding prescribed radiation exposure or release limits.

f) An observed inadequacy in the implementadon of either administrative or 1 procedural controls which could result in operation of the reactor outside the limiting condidons for operation.

g) Release of radioactivity from the site above limits specified in 10CFR20.

1.30. Rod. Control - A control rod is a device containing neutron absorbing material which is j i used to control the nuclear fission chain reaction. The control rods are coupled to the control rod drive systems in a way that allows the control rods to perform a safety j function.

l l

i l

Amendment Na 7 l 3

%~

1.31. Routine Exneriment . A routine experiment is an experiment which involves operadons I under condidons which have been extensively examined in the course of the reactor test programs and which is not defined as any other kind of experiment. Experiments and classes of experiments which are to be considered as rouune experiments must be so defined by the Reactor Operations Committee.

1.32. Safetv 1 imit - A Safety IJmit is a limit on an important process variable which is found to l be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. De principal physical barrier is the fuel element cladding.

1.33. Scram time . Scram Time is the time required to fully insert the control rods following dic l actuation of a 1.imiting Safety System Setung.

1.34. Secured Exneriment . A Secured Experiment is any experiment, experimental f ellity, or l component of an experiment that is held in a stadonary position relative to die reactor by mechanical means. De restraining forces must tv substantially greater than those to which the experiment might be subjected by hydraulic, pneumauc, buoyant, or other

, forces which are normal to the operating environment of the experiment, or by forces l which can arise as the result of credible malfunctions.

l 1.35, Shall. Should. and May he word "shall" is used to denote i requirement, the word l "should" denotes a recommendation, and the word "may" denotes permission, neither a requirement not a recommendation.

1.36. Shutdown Marcin - Shutdown Margin is the reactivity existing when the most reactive l control rod is fully withdrawn from the core and the other control rods are fully inserted into the core.

1.37. Snecial Exneriments Special experiments are experiments which are neither routine l experiments nor modilled roudne experiments.

1.38. TRIGA Fuel Element - A TRIGA fuel element is a sealed unit containing (U,Zr)ll, fuel l for the reactor. The uranium is enriched to less than 20% in 235 U and the fracdon of hydrogen is in the range of 1.0-1.1 for aluminum-clad TRIG A elements and in the range of 1.6-1.7 for stainless-steel clad TRIGA elements.

3 Amendment No 7 4.

____m

, _ . . _ - . - - - _ .__ _ _ _ m._ ._.- - .. _ _ _ _ . . _ . . - _

.. _7 _

• F 2 RAFETY I_IMITS AND 1.lMrITNG SAFETY SYS'mM SETrlNGS:

2.1, Safety i Imit (SI3 Arvilcability This specification applies to the temperature of the reactor fuel, ,

. Oblective he objective of this specification is to define the maximum fuel temperature that

can be permitted with confidence that no damage to the fuel element will result.

Snecification

- The temperature in any fuel element in the Dow TRIGA Research Reactor shall not exceed 500 C under any conditions of operation.

M A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel and the cladding if the fuel temperature exceeds the safety limit. De pressure is caused by the heating of air, fission product gases, and hydrogen from the dissociation of the fuel-moderator. T1w -

magnitude of this pressure is determined by the temperature of the fuel element and by the hydrogen content. Data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of Zrlit .6 will remain below the ultimate

stress provided that the fuel temperature does not exceed 1050 C and the fuel cladding temperature does not exceed 500 C. When the cladding temperature can equal the fuel temperature the fuel temperature design limit is 950 C (M. T.

Simnad, G.A. Project No. 4314, Report e 117 833,1980).

I Experience with operation of TRIGA fueled reactors at power levels up to 1500 kW shows no damage to the fuel due to thermally induced pressures.

The thermal characteristics of aluminum-clad TRIGA fuel elements using Zrlina moderator have been analyzed (S C llawley and R. L, Kathren, NUREG/CR 2387, PNL-4028, Credible Accident Analyses for TRIGA and TRIGA fueled Reactors,1982). A loss-of-coolant analysis showed that in a typical graphite-reflected Mark l TRIGA reactor fueled with 60 aluminum-clad fuel elements (Reed College) the maximum fuel temperature would be less than 150 C following infinite operation at 250 kilowatts terminated by the instantaneous loss of water. Rese temperatures are well below the region where the n+ 6 + Y to a

+ 6 phase change occurs in Zrlii a (560 C).

Ameadnwat No. 7 5

m 4

2.2, lJmitlne Safety System Settinyn U RRC) '

Armllenhility This speel!1 cation applies to the reactor cram setting which prevents the reactor

- fuel temperature from reachliig the safety limit, Objective The objective of this specificadon is to provide a reactor scram to prevent the safety limit from being reached, Enecifiention The LSSS shall not exceed 300 kW as measured by the calibrated power channels, DALLs The LSSS which does not exceed 300 kW provides a considerable safety margin, .

One TRIGA reactor (General Atomics, Torrey Pines) showed a maximum fuel  !

temperature of 350 C during operadon at 1500 kilowatts, while a 250-kilowatt TRIGA teactor (Reed College) showed a maximum fuel temperature of less than

= 150 C (reponed by S. C, llawley, R, L. Kathren NUREG/CR 2387, PNL-4028 (1982), Credible Accident Annivses for TRIGA nnd TRIGA-Fueled Remetnrs).' A i

portion of the safety margin could be used to account for variations of flux level (and thus the power density) at various parts of the core. The safety margin should be ample to compensate for other uncertainties, including power transients during otherwise steady state operation, and should be adequate to protect .

' aluminum clad fuel' elements from cladding failure due to temperature and pressure 1 effects, I

i l

i, Amendment No. 7 6

I

- .- . - . . - . ~ . . - _ . _ _ __ _ -.. ---..-. . _ - . - . . . - - . , - . ,. . - . , . . . - , -. .

i c l ..

i

- 3, LIMITINO CONDITIONS FOR OPERATION (LCO)

- 3,1,' Renetivity I.imits d

' Annllenhillev hese specifications shall apply to the reactor at all times that it is in operation.

4-

- Oblective 4

, he purpose of the specification is to ensure that the reactor can be controlled and shut down at all times and that the safety limit will not be exceeded.

i Eneciflentions he reactor shall be shutdown by more than $.50 with the most reactive control rod fully withdrawn, the other two control rods fully inserted, cold, no menon,

, including the reactivity worth of all experiments.

t i

he excess reactivity measured at less than 10 watts in the reference core 1: condition, with or without experiments in place, shall not be greater than $3.00.

ILuc1 he value of the minimum shut (k)wn margin assures that the reactor can be safely

shut down using only the two least reactive control rods, he assignment of a specification to the maximum excess reactivity serves as an I

additional restriction on the shutdown margin and limits the maximum power excursion that could take place in the event of failure of all of the power level safety circuits and administrative controls.

3 4

e i

I- Amendment No. 7

.7 i

.m - _-.. , _.

09

_3,2 Core Cnnfleuratinn Armlienhllity -

This specification applies to the core configuration.

Objective

- *lhe objective of this specificadon is to assure that the safety limit will not be exceeded due to power pealdng effects, Mneciflentions The critical core shall be an assembly of standard NRC appro',a,d stainless steci clad or aluminum clad TRIGA fuel elements in light water, The fuel shall be arranged in a close-packed array for operadon at full licensed power except for (1) replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes and (2) the start up neutron source.

The aluminum-clad fuel element shall be placed in the E or F ring of the core.

Rascs Operation with standard NRC approved TRIGA fuel in the standard configuradon ensures a cons:rvative limitation with respect to the Safety Limit.

Placement of the aluminum clad fuel element in the outer rings of the reactor core will help ensure that this element is not exposed to higher than average power -

levels, thus providing a greater degree of conservadsm with respect to the Safety;

- Limit for this one element.

Amendment No 7 3-

i e

3.3. Reactor Control and Safety Systems Annlicability hese specifications apply to the reactor conuol and safety systems and safety related instrumentation that must be operating when the reactor is in operation.

Objective De objective of these specifcations is to assure that all reactor control and safety systems and safety related instrumentation are operable to minimum acceptable standards during operation of the reactor.

Soccincations here shall be a minimum of one scram-capable analog safety channel.

Dere shall be a minimum of three operable control rods in the reactor core.

Each of the three control rods shall drop from the fully withdrawn position to the fully inserted position in a time not to exceed one second.

We reactor safety channels and the interlocks shall be operable in accordance with table 3.3A.

De reactor shall not be operated unless the measuring channels l'sted in Table 3.3B are operable.

Positive reactivity insertion rate by control rod motion shall not exceed $.20 per second.

Amendment No, 7 9

.x-------------_----_-----______. ---_-- - - _ _ _ __ _ - .

G-Daici Safety channels with scram capability udlir.ing analog circuitry have been proven acceptable by more dian thirty years of experience, i ne requirement for three operable control rods ensures that the reactor can meet the shutdown specifications.

De control rod drop time specification assures that the reactor can be shutdown -

promptly when a scram signal is initiated. We value of the control rod drop time is adequate to assure safety of the reactor.

l Use of the specilled reactor safety channels, set points, and interlocks given in table 3.3A assures protection against operation of the reactor outside the safety limits.

De requirement for the specified measurement circuits provides assurance that important reactor operation parameters can be monitored during operation.

De specification of maximum positive reactivity insertion rate helps assure that the Safety Limit is not exceeded.

Amendment No.7 10 .

.y e

TABLE 3.3A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Chanrw l Minimum Oncrable Scrarn Settwilnt ..

Reactor Power Level 2 Not to exceed maximum licensed NM1000 & NPP1000 power NPP1000 1 - Failure of the detector Detector liigh Voltage Power Supply high voltage power supply-NM1000 1 Fallure of the detector Detector High Voltage Power Supply high-voltage power supply .

Manual Scram 1 - Not applicable Watchdog (DAC to CSC) i Not applicable Inter:ocks InterlockK'hannel Function Startup Countrate Prevent control rod withdrawal when the neutron count rate is less than 2 cps Rod Drive Control Prevent simultaneous manual withdrawal of two control elements by the control rod drive motors Reactor Period Prevent control rod withdrawal when the petiod is -

less than 3 seconds 1

4 l Amendment No. 7

, I i

y

('

TABLE 3.3A 1

t BASES FOR REACTOR SAFETY CHANNELS AND INTERLOCKS Scram Channels . q l

Scram Phnnnel g Reactor Power Level Provides ' assurance that the reactor will be shut down . l

> automatically before the safety limit can be exceeded I

Reactor Power Channel Provides assurance that the reactor, Detector Power Supplies cannot be operated without powcr to the neutron detectors which provide loput to tLe NM1000 and NPP1000 power channels .

Manual Scram - Allows the operator to shut the reactor doim at any ,

indication of unsafe or abnormal conditions -

Watchdog Ensures adequate communications between the Data Acquisition Computer (DAC) and the Control System Computer (CSC) units.

Interiocks Interlock / Channel h Startup Countrate Provides assurance that the signal in the NM1000 channel I is ade_quate to allow reliable indication of the state of the.

neutron chain reaction.

Rod Drive Control . Limits the maximum positive reactivity insertion rate Reactor Perkx! Prevents operation in a regime in which transients could cause the limiting safety system setting to be exceeded Amendment No. 7 L2 -

v

.e TABLE 3.3D MEASURING CHANNELS hicasuring channel hiinimum Number Operable Nh11000 1 NPP1000 1 Water Radioactivity I hionitor Water Temperature I hionitor TABLE 3.3P DASES FOR h1EASURING CHANNELS Measuring Channel M Nh11000 Provides assurance that the reactor power level can be l adequately monitored.

NPP10(X) Provides assurance that the reactor power level can be l adequately monitored.

Water Radioactivity Provides assurance that the water hionitor radioactivity level can be adequately monitored.

Water Temperature Provides assurance that the water hionitor temperature can be adequately monitored.

Amra *:nent No. 7 13 -

e. . _

.. i 3,4, Cmlant Svasem .j Annlienhility -

These specifications apply to the quality of the coolant in contact with the fuel cladding, to the level of the coolant in the pool, and to the bulk temperature of the coolant.  ;

Ohlectives De objectives of this specificadon are: .[

I to minimize corrosion of the cladding of the fuel elements and minimize neutron

. activadon of dissolved materials,  ;

to detect releases of radioactive materials to the coolant before such releases :

become significant, to ensure the presence of an adequate quantity of cooling and shiekling water in 'F the pool, and . i to prevent thermal degradation of the ion exct.ange resin in the purification system.

Sneciftentions i

De conductivity of the pool water shall not exceed 5 pmhos/cm averaged over one

- month.--

De pool water pli shall be in the range of 4 to 7.5.

He amount of radioactivity in the pool water shall not exceed 0,1 pCi/mL

. De water must cover the core of the reactor to a minimum depth of 15 feet during ,

operation of the reactor.

De bulk temperature of the coolant shall not exceed 60 C during operation of the reactor.

'I Amendment No, 7

+ 84 m.,. - ._, #. _ . ~ . -

y e

Dases increased levels of conductivity in aqueous systems indicate the presence of corrosion products and promote more corrosion. Experience with water quality control at many reactor facilities, including operadon of the Dow TRIGA Research Reactor since 1967, has shown that maintenance widdn the specified limit I provides acceptable control. hiaintaining low levels of dissolved electrolytes in the pool water also reduces the amount of induced radioactivity, in turn decreasing the exposure of personnel to ionizing radiadon during operadon and maintenance.

Both of these results are in accordance with the ALARA program.

hionitoring the pil of the pool water provides early detecdon of extreme values of pil which could enhance corrosion.

hionitoring the radioactivity in the pool water serves to provide early detection of possible cladding failures. Limitation of the radioacdvity according to this specification decreases the exposure of personnel to ionizing radiation during operation and maintenance in accordance with the ALARA program.

hiaintaining the specilled depth of water in the pool provides shielding of die radioactive core widch reduces the exposure of personnel to ionizing radiadon in accordance with the ALARA program, hiaintaining the bulk temperature of the coolant below the speellied limit assures minimal thermal degradation of the ion exchange resin.

Ameadment No 7 15 l

1

.m.

4 s 3.5, Confinement Annilenhltify This specincation applies to the reactor room conf r, ment.

Objective he objecuve of this specification is to midgate the consequences of possible

. release of radioactive materials to unrestricted areas.

Knecifienflon The vendlation system shall be operable and the external door (Door 10) shall be l closed whenever the reactor is operated, fuel is manipulated, or radioactive -

-: materials with the potential of airborne releases are handled in the reactor room, llails his specincadon ensures that the connnement is configured to control any -

releases of radioactive material during fuel handling, reactor operation, or the handling of possible althorne radioactive material in the reactor room.

Amendmest No. 7 i

.a.

l

e 3.6. EndIntion Monitorine systems Armlicability nese speciHcations apply to the radiadon monitoring information available to the reactor operator during operadon of the reactor.

. Oblective The objective of these specifications is to ensure that the reactor operator has adequate information to anure safe op' ration of the reactor.

Socci'ications A Continuous Air Monitor (CAM) (with readout meter and audible alarm) to measure radioactive particulates in the reactor room must be operaung during l operation of the reactor.

The Area Monitor (AM)(with readout meter and audible alarm) in the reactor room must be operating during operation of the reactor or when work is being done on or around the reactor core or experimental facilities, During short periods of repair to this monitor, not to exceed thirty days, reactor operations or work on or around the core or experimental facilities may continue while a portable gamma sensitive ion chamber is utilized as a temporary subsdtute, provided that the substitute can be monitored by the reactor operator, e Ilaics he radiation monitors provide information of existing levels of radiation and air borne radioactive materials which could endanger operating personnel or which could warn of possible inalfunctions of the reactor or the experiments in the reactor.

Amendment No. 7 17 -

e 3,7 &neriments Apnficability These specifications apply to experiments installed in the reactor and its experimental facilides.

Objective "Ihe objecdve of these specifications is to prevent damage to the reactor or excessive release et radioactive materials in case of failure of an experiment.

Soccifications

1. Operadon of the reactor for any purpose shall require the review and approval of the appropriate persons or groups of persons, except that operation of the reactor for the purpose of performing routine checkouts, where written procedures exist for those operations, shall be authMzed by the written procedures. An operation shall not be approved unless the evaluation allows the conclusion that the failure of an experiment will not lead to the direct failure of a fuel element or of any other experiment.

2. *Ihe total absolute reactivity worth of in core experiments shall not exceed

$1.00. This includes the potential reactivity which might result from experimental malfunction, experiment ikxxiing or voiding, or the removal or insertion of experiments.

3. Experiments having reactivity worths of greater than $0.75 shall be securely located or fastened to prevent inadvertent movement during reactor operation.

4. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potendally explosive materials or liquid fissionable materials shall be doubly encapsulated.

5. Materials which could react in a way which could damage the components of the reactor (such as gunpowder, dynamite TNT, nitroglycerin, or PETN) shall not be irradiated in quan' ides greater than 25 milligrams in the reactor or experimental facilities without out-of-core tests which shall indicate that, with the containment provided, no damage to the reactor or its components shall occur upon reaction. Such materials in quantides less than 25 milligrams may be irradiated provided that the pressure produced in the expertinent container upon reaction shall be calculated and/or experimentally demonstrated to be less than the design pressure of the container. Such materials must be packaged ir. the appropriate containers before being brought into the reactor room or must be in the custody of duly authoitzed hical, state, or federal officers.

Amendment No 7 18 -

6. Experiment materints, except fuel materials, which could off-gas, sublime, volatillic or produce aerosols under (a) normal operaung conaldons of de experiment or the reactor (b) credibic accident condidons in the reactor or (c) possible accident condidons in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity would not exceed the limits of Appendix B of 10 CFR Part 20.

De following assumptions should be used in calculations regarding experiments:

a. if the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation levels, the l assumpdon shall be used that 10% of the gaseous acdvity or i

aerosols produced will escape, j b. If the effluent from an . erimental facility exhausts through a litter installadon designed for greater than 99% etliciency for 0.3 micron particles, the assumpdon shall be used that 10% of the aerosols produced eschpe, c, For materials whow boiling point is above 55 C and where vapors formed by boiling this material could escape on;y through an undisturbed column of water above the core, the assumption shall be used that 10% of these vapors escape.

7. Each fueled experiment shall be controlled such that the total inventory of h) dine isotopes 131 through 135 in the experiment is no greater than 1.5 curies and the maximum strontium 90 inventory is no greater than 5 millicuries.

8. If an experiment container fails and releases material which could damage the reactor fuel or structure by corTosion or other means, physical inspection shall be performed to determine the consequences and the need for corrective acdon.

9. Experiments shall not occupy adjacent fuel element posidons in the B and C rings.

Amendmeal No 7 19 -

1

Bases

l. This specificadon is intended to provide at icast one level of review of any proposed operation of the reactor in order to minimize the possibility of operations of the reactor wtJch could be dangerous or in violadon of administradve procedures or the technical specifications. The excepdon is made in the case of those few very well characterized opuadons which are necessary for routine checkout of the reactor and its systems, provided that those operations have been defined by written procedures which have been reviewed and approved by the Reactor Supervisor and the Reactor Operations Committee.

2. This specificadon is intended to limit the reacuvity of the system so that the Safety Limit would not be exceeded even if the contribution to the total reactivity by the experiment reactivity should be suddenly removed. l

3. This specification is latended to Ilmit the power excursions which might be induced by the changes in reacdvity due to inadvertent motion of an unsecured experiment, huch excursions could lead to an inability to control the reactor within the limits imposed by the lleense.

d

4. Tids specification is intended to reduce the possibility of damage to the ret.ctor or the experiments due to release of the listed materials.

5. ' Tids specification'is intended to reduce the possibility of damage to the reactor in case of accidental detonation of the listed materials.

6. This specification is intended to reduce the severity of the results of acciden al release of airborne radioactive materials to the reactor room or the atmosphere.

7. This specilication is intended to reduce the severity of any possible release of these. fission products which pose the greatest hazard to workers and the general public.

8. This specl0 cation requires specific actions to determine the extent of damage folicwing releases of materials. No theoretical calculadons or etaluadons are allowed.

[ 9. Tids specification prevents serious modification of the neutron distribudon which could affect the ability of the control rods to perform their intencted function of malataining safe control of the reactor, u

i k

Ameaament No. 7

. :o .

4. SURVEll I ANCE REOUIREMENTS Allowable surveillance intervals shall not exceed the following:

bientdally not to exceed 30 months annually not to exceed 15 months semi annually not to exceed seven and one-half months  ;

quarterly - not to exceed four months I monthly not to exceed six weeks weekly not to exceed 10 days daily must be done before the commencement of operation each day of operation Established frequencies shall be maintained over the long term, so, for example, any monthly surveillance shall be performed at least 12 times during a calendar year of normal operation. If the y reactor is not operated for a period of Ome exceeding any required surveillance interval, that surveillance task shall be performed before the next operation of the reactor. Any surveillance tasks which are missed more than once during such a shut-down interval need be performed only once before operation of the reactor. Surveillance tasks scheduled daily or weekly which cannot be performed while the reactor is operating may be pastponed during continuous operation of the reactor over extended times. Such postponed tasks shall be performed following shutdown after the extended period of continuous operation before any further operation, where each task shall be performed only once no matter how many times that task has been postponed.

Amentiment No 7 21 -

1 l

g -.

.i,

- 4,liMr Core Parametern -

Applicability Dese specifications apply to' surveillance requirements for reactor core parameters.

Ohlective -

De objective of these specifications is 'to ensure that the specifications of section 3.1 are satisfied.

I Rnecification The reactivity worth of each control rod, the reactor core excess, and the reactor shutdown margin shall be measured at least annually and after each time the core i fuelis moved, 111111 Movement of the core fuel could change the reactivity of the core and thus affect both the core excess reactivity and the shutdown margin, as well as affecting the -

worth of the individual control rods. Evaluation of these parameters is therefore -

. required after any such movement.' Without any such movement the changes of

- these parameters over an extended period of time and operation of the reactor:

.have been shown to be very small so that an annual measurement is sufficient to ensure compliance with the specifications of section 3.1.

s Amendment No. 7 22

i 4

4,2. Remetor Cnntrol and Rafety Systems Arr311cnhility

'Ihese speci0 cations apply to the surveillance requirements of the reactor safety systems.

Oblective The objective of these specifications is to ensure the operability of the reactor 1 safety systems as described in section 3.3.

Snecincations

1. Control rod drive withdrawal speeds and control rod drop times shall be measured at least annually and whenever maintenance is performed or repairs are made that could affect the rods or control rod drives.

2. A channel calibration shall be performed for the NM1000 power level l channel by thermal power calibration at least annually.

3. A channel test shall be performed at least daily and af'er any mentenance or repair for each of the six scram channels and each of the three interlocks

- listed in table 3.3A.

4. 'The control rods shall be visually inspected at least biennially.

Bases

1. Measurement of the control rod drop time and compliance with the __

specification indicates that the control rods can perform the safety function properly. Measurement of the control rod withdrawal speed ensures that the maximum reactivity addition rate specification will not be exceeded, g,

2. Variations of the indicated power level due to minor variations of either of the two neutron detectors would be readily evident during day to-day operation. The specification for thermai 'alibration of the NM1000 channel provioes assurance that long term drift of both neutron detectors would be detected and that the reactor will be operated within the authorized power range.

1'

3. The cham.J tests performed daily oefore operation and after any repair or maintenance provide timely assurance that the systems will operate properly during operation of the reactor.

4. Visual inspection of the control rods provides opportunity to evaluate any corrosion. distortion, or damage that might occur in time to avoid malfunction of the control rods. Experience at the Dow TRIGA Reactor Facility since 1967 indicates that the surveillance srcification is adequate to assure l proper operation of the control rods. 'thu surveillance complements the rod drop time measurements.

Amendment No. 7

.n.

--m - -

m

=4

-- 4.3. Conlant Svttem -

Annllenhility These specifications shall apply to the surveillance tequirements for the reactor coolant system. _.

. Ohlective The objective of these specificadons is to ensure that the specificadons of section:

3.4 are satisfied.

Sneciflentinns

1. The conductivity, pH, and the radioactivity of the pool water shall De

. measured at least monthly.

~

2. . The level of the water in the pool shall be determined to be adequate on a weekly basis.

3. l'ic temperature of the coolant shall be monitored during operadon of the reactor.

ILuci

1. Experience at the Dow TRIGA Research Reactor shows that this specification is adequate to detect the onset of degradation of the quality of the pool water in a timely fashion. Evaluation of the radioactivity in the pool .

< ' water allows the detection of fission product releases from damaged fuel elements or damaged experiments.

2. Experience indicates that this specification is adequate to detect losses of pool water by evaporation.

- 3. This specificadon will enable operators to take appropriate action when the -

coolant temperature approaches the specified limit.

f l

i 1

Amendmsat No. 7 24

. . . ~ . - . - -, . _ _ . . . , . . - , __ - .

y.

4,4, Radintion Monitorine Systems

~

t A_inflcabilltv These specificsons apply to the surveillance requirements for the Continuous Air Monitor (CAM) and the Area Monitor (AM), both locatea in the reactor room.

l . Obiective l -- *Ihe objective of these specifications is to ensure the quality of the data presented by these two instruments, Snecifications

1. A channel calibration shall be made for the CAM and the AM at least annually.

2. A channel test shall be made for the CAM and the AM at least weekly.

llaics "Ihese specifications cr sure that the named equipment can perform the required functions when the reactor is operating and that deterioration of the instruments will be detected in a timely manner. Experience with these instruments has -

shown that the surveillance intervals are adequate to provide the required assurance, s

)

l Ameedment No 7 4 25

a-4.5. Facility Snecluc Surveillance Arvilenbilltv

'This specification shall apply to the fuel elements of the Dow TRIGA Research Reactor.

Oblective The objective of this specification is to ensute that the reactor is not operated with damaged fuel elements.

Snecl0 cation Each fuel element shall be examined visually and for changes in transverse bend and length at least once each five years, with at least 20 percent of the fuel elements examined each year. If a damaged fuel element is Identified, the entire inventory of fuel elements will be inspected prior to subsequent operations.

'The reactor sha9 not be piernted with damaged fuel except to detect and identify damaged fuel for removal. A TRIGA fuel element shall be considered damaged l l and removed from the core if:

l a) "Ihe transverse bend exceeds 0.125 inch over the length of the cladding.

b) 'lhe length exceeds the original length by 0.125 inch.

c) A clad defect exists as indicated by release of fission products.

Dasis Visual examination of the fuel elements allows early detection of signs of deterioratJon of the fuel elements, indicated by signs of changes of corrosion patterns or of swelling, bending, or elongation. Experience at the Dow TRIGA Research reactor and at other TRIGA reactors indicates that examination of a live-year cycle is adequate to detect problems, especially in TRIGA reactors that

, are not pulsed. A live-year cycle reduces the handling of the fuel elements and thus reduces the risk of accident or damage due to handling.

Aniendment No. 7

= 16 s

4 4.6. ALMLA Applicabilitv

. This speellication applies to the surveillance of all reactor operadons that could result in occupational exposures to ionizing radiation or the release of radioactive materials to the environmer.t.

Ohlective The objective of this specification is to provide surveillance of all operations that could lead to occupational exposures to ionizing radiation or the release of radioactive materials to the environs.

Rnecincation The review of all operations shall include consideration of reasonable alternate operational modes which might reduce exposures to ionizing radiation or releases of radioactive materials.

Dashi

, Experience has shown that experiments and operational requirements, in many .

cases, may be satisfled with a variety of combinations of facility options, power levels, time delays, and effluent or staff radiation exposures. The ALARA (As .

Low As Reasonably Achievable) principle shall be a part of overall reactor operation and deLiled experiment planning.

Amendment No. 7

. 27

.y 5, DESIGN FEATURES 5.1. Henctor Rite and Bull <tino Amlicabil)!y, These specifications shall apply to the Dow TRIGA Research Reactor.

Obiectives The objectives of these specifications are to define the exclusion area and characteristics of the confinement.

Snecifications The minimum distance from the center of the reactor pool to the boundary of the exclusion area shall be 75 feet.

The reactor shall be housed in a room of about 6000 cubic feet volume designed to restrict leakage.

1 All air or other gas exhausted from the reactor room and from associated experimental facilities during reactor operation shall be released to the environment at a minimum of 8 feet above ground level. ,

DECS The minimum distance from the pool to the boundary provides for dilution of effluents and for control of access to the reactor area.

i Restriction ofleakage, in the event of a release of radioactive materials, can contain the materials and reduce exposure of the public. ,

Release o^ bases at a minimum height of 8 feet reduces the possibility of exposure of perscutel to such gases, i

i Amesdment No. 7

.a.

l

ee 5.2. llenctor Coolant Evstern App!!cability This specincation app!!cs to the Dow TRIGA Research Reactor.

DNCCthe lhe objective of this $recincation is to define the charactettstics of the cooling systern of this reactor.

StecInc:11 ton I

The teactor core shall be coolal by r.atural convective water flow.

lhih lixperience has shown that TRIGA reactors operating at power levels up to 1(XX) kilowatts can be coolett by natural convective water flow without damage of the fuel clernents, i

Akadunt Ni> 1

' 2V -

m j

• i 5.3. newmr enre ami Fuel Armileshilltv 1hese specifications shall be applicable to the Dow TRIGA Research Reactor.

Oblective "Ihe objective of these specificadons is to define certain characterisucs of the reactor in order to assure that the design and accident analyses will be correct.

heincation The fuel will be standard NRC approved TRIGA fuel.

'ihe conuol elements shall have tcram capability and shall contain tersted graphite, boron carbide powder, or boron and its components in solid form as a poison in an aluminum or stainless steel clackling.

The reflector (excluding experiments and experimental facilities) shall be a combination of graphite and water.

Ilaics

'ihe entire design and accident analysis is based upon the characteristics of TRIGA fuel. Any other material would invalidate the findings of these analyscs.

'Ihc control elements perform their function through the absorption of neutrons, thus affecting the reactivity of the system._ Iloron has been found to be a stable and effective ma:erial for this control.

• lhe tellector serves to conserve neutrons and to reduce the amount of fuel that must be in t!.e core to maintain the chain reaction.

Amendment No. 7 30

t to 5.4.1:uel storge Ann 11cnbility This specllication applies to the Dow TRIGA Research Reactor fuel storage facilities.

Objective lhe objective of this specification is the safe storage of fuel.

Snecification All fuel and fueled devices not in the core of the reactor shall be stored in such a way that k,r shall be less than 0.8 under all conditions of in(xleration, and that will perrnit sullicient cooling by natural convection of *vates or air that ternperatures shall not exceed the Safety Lirnit.

IIAih A value of k,rr of less than 0.8 precludes any possibility of inadvertent establishiner.t of a self sustaining nuclear chain reaction, Cooling which inalntains teinperatures lower than the Safety IJinit prevents possible darnage to the devices with subsequent release of radioactive inaterials.

ANW9dHW81 No 7 3)

6. Al%11NISTRATIVE CONTROLS 6.l. Organization The Dow TRIGA Research Reactor is owned and operated by The Dow Chemical Company.  :

The reactor is administered and operated through the Analytical Sciences Laboratory of the l >

hilchigan Division of Dow Chemical USA and is located in 1602 Building of the Analytical Sciences Laboratory at the hiidland, hiichigan location of the hiichigan Division. l 6.l.1. Structure The structure of the administration of the reactor is shown in figure 6.1. This structure cuts across the lines of management of The Dow Chemical Company.

The individual responsible for radiation safety is the Radiation Safety Officer for the reactor who reports on matters of radiation safety to the Radiation Sakty Committee and to the Reactor Operations Committee. The Radiation Safety committee oversees the radiation safety program and is responsible for its irnplementation. The review and addit functions are performed by the Reactor Operations Committee which is composed of at least four persons including a manager within Analytical Sciences l Laboratory, the Radiation Safety Officer. and the Reactor Supervisor.

6.1.2. Responsibility The day to-day responsibility for the safe operation of the reactor rests with the Reactor Supervisor who is a licensed Senior Reactor Operator appointed by the <

Facility Director or manager within Analytical Sciences Laboratory. The Reactor l Supervisor may appoint equally qualified individuals, upon notification of the Facility Director and the Reactor Operations Committee, to assume the responsibilities of the Reactor Supervisor. The Reactor Supervisor reports in a management sense to the Facility Director and within the reactor organization to the

Reactor Operations Committee.

j 1

i 6

i i

d I

Anendnwnt No 7 32

V Figure 6.1. Administration l

hinnager, Industrial llyriene Research and Radiation Safety Senior Research 4 hianager Technology Comrnittee Chair, RSC Chair R.O.C.

AA 4 I V I I

_J -+ Facility VVV Director y i I Reactor Operations Supervisor, l Committee (ROC) y Industrial l Ilygiene l $ h

__ y Reactor Supervisor l l I

1 I

i i-+

V I l l y M l i Radiation Safety Officer 4__J l 1.icensed SROs 4-__ and ROs Line inanagernent responsibilities i

Line htanagement Reporting i --- Communication Reponing l

Arrendirent No 7 1).

6e 6.l.3. Staffing 1he niittimum staffing when the reactor is not secured shall be:

a a licensed Reactor Operator or Senior Reactor Operator in the control roorn, and

b. a second person present at the facility able to carry out prescribed written .

Instructions, and c, a licensed Senior Reactor Operator in the facility or readily available on call and able to be at the facility within 30 minutes.

The following operations require the presence of the Reactor Supervisor or a designated alternale:

a. manipulations of fuel in the core;

b. manual removal of control rods; e, inalntenance performed on the core or the control rods;

d. recovery !!om unexplained heramt, and

c. movement of any in core experiment having an estimated reactivity value greater than 50.75.

A list of reactor facility [rtsonnel by name and telephone number shall be readily available in the control room for use by the operator, including management, radiation safety, and other operations personnel.

6.1.4. Selection and Training of Personnel The Reactor Supervisor is responsible for the training and requalification of the facility Reactor Operators and Senior Reactor Operators.

The sciection, training, and requalification of operations personnel shall be consistent with all current regulations.

Day to-day changes in equipment, procedures, and r.peellications shall be communicated to the facility staf f as the changes occur.

J Amendment No 7 34 .

r O

6.2. Review and Audit ,

The review and audit functions shall be the responsibility of the Reactor Operations  !

Committee (ROC).

6.2.1. Charter and Rules

a. This committec shall consist of Facility Director, who $1 pil be designated the chair of this committee; the Radiation Safety Officer; the Reactor 51pervisor; and one or more persons who are competent in the field of reactor operations, radiation science, or reactor / radiation engineering. In the event that the positions of theReactor Supervisor and Facility Director are filled by the same person, a manger within Analytical Sciences Laboratory shall serve as the chair of the committee,

b. A quomm shall consist of a majority of the members of the ROC. No more than one half of the voting members present shall be members of the day to-day reactor operating staff.

c. The Committee shall meet quarterly and as often as required to transact ,

business,

d. Minutes of the meetings shall be kept as records for the facility,

c. In cases where quick action is necessary members of the ROC may be polled by -

telephone for guidance and approvals.

f. The ROC Shall report at least twice per year to the Radiation Safety Committee.

t Anwndnw n No. 7 35

l l

I l

6.2.2. Review Functions The ROC shall review and approve:

a. every experiment involving fissionable material; b experiments or operations which would require a change of core condguration, or a change in the equipment or apparatus associated with the reactor core or its irradiation facilities, or a new piece of apparatus being mounted in the reactor well; except that movement of the neutron source for the purpose of routinely checking the instrumentation, or the movement of the neutron detectors to establish the proper calibration of the associated channels shall not require review by the ROC;

c. any other experiment or operation which is of a type not previously approved by the Committee;

d. proposed changes in operating procedures, technical specifications, license, or charter;

c. violations of technical speellications, of the license, of internal procedures, and of instructions having safety significance;

f. operating abnormalities having safety significance;

g. reportable occurrences; h, proposed changes in equipment, systems, tests, or experiments with respect to unreviewed safety questions; and

1. audit reports.

i l

l Altienduwst No. 7 l

36 .

6.2.3. Audit Function

a. The ROC shall direct an annual audit of the facility operations for conformance to the technical specillcations, license, and operating procedures, and for the results of actions taken to correct those deficiencies which may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety.

1hlt audit may consist of examinations of any facility tecords, review of procedures, and interviews oflicensed Reactor Operators and Senior Reactor Operators.

The audit shall be performed by one or more persons appointed by die ROC. At

-least one of the auditors shall be famillar with reactor operations. No person directly resp (msible for any portion of the operation of the facility shall audit that operation.

A written report of the audit shall be submitted to the ROC within three months of the audit.

Dellclencies that af fect reactor safeif shall be reported to the Facility Director immediately,

b. The ROC shall direct an annual audit of the facility emergency plan, security plan, and the reactor operator requalillcation program. This audit may consist of the annual review of these plans for the requalllication program.

Amendment No 7

• 37 i

J 6.3 Procedures Written procedures shall be reviewed and approved by the ROC for:

s. reactor startup, routine operation, and shutdown; b, ernergency and abnortnal operating events, including shutdown;

c. fuelloading or unloadjng;

d. control rod temoval or installation;

c. checkout, calibration and deletmination of operability of reactor operating instrumentation and controls, control rod drives and area radiation and air paniculate inonitors; and

f. preventive maintenance procedurcq.

Temporary deviations from the procedures may be rnade by the responsible Sertior Reactor Operator or higher individual in order to deal with special or unusual circumstances. Such deviations shall be documented and reported Irmnediately to the Reactor Operations Committee.

6.4. Etneriment Review and Aoproval

a. Routine Experiments (as reviewed and defined by the ROC) shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor,

b. Modithd Routine Experiments shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor. The written approval shall include documentation that the hazards have been considend by the leviewer and tren found appropriate for this form of experiment,

c. Special Experiments, those experiments that are neither Routine Experiments nor Modified Routine Experiments, shall have the approval of toth the Reactor Supervisor (or designated Assistant Reactor Supenisor) and the ROC.

Experiments which require the approval of the ROC through sections 6.2.2.a..

6.2.2.b., or 6.2.2.e. of the Technical Speellications are always Special Exgeriments.

Anwedawat No 7 31

6.5, Required Actiont 6.5.1, in case of Safety Lirnit violation:

a. the reactor shall be shut down until resumed operations are authorized by the US NRC; and I

b. the Safety Limit violation shall be linmediately reported to the Facilhy Director or to a higher level; and l

c. the Safety Lirnit violation shall be reported to the US NRC in accordance with section 6.6.2.; and

d. a report shall be prepared for the ROC describing the applicable circurnstances leading to the violation including, when known the cause and contributing factors, describing the effect of the vlulation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public, and describing corrective action taken to prevent recurrence of the violt.uon 6.5.2. In e.se of a Reportable Occunence of the type Identitled in section 1.29 l

a. reactor conditions shall be returned to normal or the reactor shall be shut down; if the reactor is shut down operation shall not be resumed unless authorized by the Facility Director or designated alternate; and l

b. the occurrence shall be reported to the Facility Director and to the US NRC as required per section 6.6.2.t and

c. the occurrence shall be reviewed by the ROC at the next scheduled meedng.

I Amendment No. 7 l

H+

l-

& nN Natupeamy

• t

'OC ll:13 01 ul papuatutuoaas Jo pamolle asotn jo <6 gt ucy: Jaica23 aJe sainsodxa yins aJays sjo Isp pue puuosJad Xilipc] Aq parpaaJ saJnsodxa uopcipei atn jo XJetutuns n (3 pun :(papaau si laajia spp on nuatuaicis e Xt uo 'papuatutuosas Jo pamolle uopenuaauoa 04:jo sSt ucy ssat si uoisnjjip Jo uopnsp Jaye ascalai a3ciare paintupsa atti ji :1uanula ay1 ut nuasaJd sapipnuoipeJ tenppipui jo oletupsa un olycapac1d luaixa 0y on apnput titys XJetutuns atp) a3Jcyasip Jo ascataJ Lpns jo sups! ay ajojaq Jo le paulunaiap se Joiciado Jaumo oy Jo [ojiuoa aApaajja ayi puoXaq suospua on pa32cyasip Jo pascaps sguantua aApacolpc1 jo lunotun pun aJnicu ay1 jo XJctutuns c 0 taaucaulu3 1s Xiajes Supey suopesado aaucuaiulutu arpaanoa pun arguaraJd Jo[cin jo uopctnqui (a taauaianaaJ apnpaid on uanci uopan arpaanoa

'a[qcattdde ajays 'Sulpnpui tuain Joj suoscal put suscpings patnpayasun ata (p ta[qualldde sum (c) 69'0911:10 01 ic41 p"~; pa^toaul asas suoissanh Xiajes pamapasun ou lets suoisnpuoa ayi on 3uipcal sasX icue ay1 jo XJctutuns c Huipnput

'podati sisXtcuy Xiales ay1 u; paqqasap lou ate pun 41snopajd patuJojiad osoyt tuoJJ iuajajlip Xilucaulu31s oJe icip stuatupadra put sisat mau jo uopcinqui pun 'sainpaaoJd pue Xilipuj Jopeas ay1 ut sa3uctp Jo[etu jo uopegnqct (a tuopciado jo sXep 1: cec 3auf ,

1c101 atu 3uipnpur 'aaaapadxa SupeJado JonacaJ Jo XJetutuns aAncucu e (y

!3ulutusi pun 'sasuaa;[ 'Jicis Xilipej atu jo snicis (e

3ulmoltoj atp apnput ticus yajym 'g[1 uoj3a>{ 311N sg 'JoicJniupupy l Icuoi 3a>{ atp ol Adoa e tpla '30 'uon3uBDcM '311N S0 MSad 1011uoa luatunaoG atu, os put aaliliutuo3 Xiains uopcipelt arp on 'sasci [cnuuc jo sauctujoj1ad 1661 Jauenh is19 agi ylla Sulucis 'Xilenuun panpugns aq Itcys podaJ y spoda}1 SupeJado . 't99 spodau 9 9 ,

b 99 l\ __. ___ ___

o r I

6.6.2. Special Reports l

a.1here shall be a report to US NRC Region ill not later than the following j working d'/ by telephone and conllrmed in writing by telegraph or similar coeveyance ta lhe Document Control Der.k. US NRC, with a copy to the Regional l Adm:nistrator, Region ill, US NRC to be followed by a written report that descrites the eve it witidn 14 days of:

l a violation of the Safety Limit: or a reportablo occultence (section 1.29).  ;

b.1here shall be a written report presented witidn 30 days to '!he Document Control  !

Desk US NRC, with a copy to the Regional Administrator, Region 111, US NRC, of' permanent changes in the facility staf f involving the reactor supervisor or the l facility director; or significant chan;es h the transient or accident analyt,ls report as described in the Safety Analysis lhevrt.  ;

c. A written remut shall be submitted to The Document Control Desk, US NRC, with ,

a copy to the Regional Administrator, Region lit, US NRC, within 60 days after t criticality of the reactor under conditions of a new facility license authorizing an increase in reactor power level, describing the measured values of the operating ,

conditions or characteristics of the reactor under the new conditions.

l s

f j.

l P

I 6

4..

Amendnwat .h 7

.n.

, , - ,-- A.-_ -~..,.- a-..-_m _.. --~ # . . - . . . _ , - . - - - - - -.-.,G,-.J . . _ . . , ---. . - - .,. . , . . _ - . - _ _ . - _ . - . . . ~ . ~ _ , - -

I 6.7. Reccids 6.7.1. 1he following records shall be kept for a rninlinum period of five years:

a. reactor operating logs;

b. Irradiation reques.t sheets;

c. checkout sheets;

d. Inanntenance records;

c. calibration records; I accolds of reportable occurrences;

g. fuel inventories, recclpts, and shipments;

h. Ininutes of ROC rueetings; L records of audits:

1 facility radiation and contamination surveys; and

k. sntveillance activilles as required by the Technical Specifications.

6.7.2 Records of the retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one complete requalification schedule.

6.7.3. The following records shall be retained for the lifetime of the reactor:

a. records of gaseous and liquid radioactive effluents released to the environment;

b. records of the radiation exposure of all individuals monitored; and

c. drawings or the reactor facility.

Anwadttwat No 7 42 t

• - - . . . ,,