ML20108A476
| ML20108A476 | |
| Person / Time | |
|---|---|
| Site: | North Carolina State University |
| Issue date: | 10/15/1984 |
| From: | Poulton B North Carolina State University, RALEIGH, NC |
| To: | Thomas C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8411150003 | |
| Download: ML20108A476 (39) | |
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$% ' North Carolina StateUniversity
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'. Department of Nuclear Engineering 15 October 1984 Box pfg Zip 2;4igypyy k
! Director, Office of Nuclear Reactor Regulation Attention: Cecil 0. Thomas, Chief Standardization and Special Projects Branch U. S. lOclear Regulatory Commission Washington, D. C. 20555 Docket No. 50-297 Facility License No. R-120 Ref: 26 July 1984 Request for Technical Specifications Chanaes.
, Dear Sire Per discussion with Mr. Robert Carter on 19 Sepember 1984, we are resubmitting the entire 26 July 1984 enclosure with minor editorial changes.
This request replaces and supercedes the previous request of 26 July 1984 and thus submitted conforms to all current Facility and USNRC procedu-ral guides.
Sincerely, b%ce Bruce R. Poulton k.h W & }7nu) /
Chancellor BRP:1pe i
Attachment:
Resubmittal of 26 July 1984 request with editorial changes.
cci (all with attachment)
USNRC, Region II Dr. K. V. Mani Dr. P. J. Turinsky Dr. J. A. Mulholland Dr. B. W. Wehring Dr. J. J. Wortman l Mr. G. D. Miller Mr. D. W. Morgan, Jr.
Subscribed and sworn to before me this /6 day of (Oefv/>vt, , 1984.
Notary Public N- &
My commission expires bel 8 /426
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8411150003 841015 r DR ADOCK 0500029 North Carchna State University is North Carolina's originalland grant institution
, I and in a constituent institution of The University of North Carolina.
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i North Carolina StateUniversity . y mm w Box 7108, Raleigh, NC 27695-7108 NJe Telepime (sio) nf. s o4 26 July 1984 sy David Clark Labs PR-100-9 6
Mr. Thomas C. Bray Nuclear Reactor Program Nuclear Engineering Dept.
Box 7909 NCSU Campus
Dear Mr. Bray:
As you are aware, a joint subcommittee of the Reactor Safeguards Advisory Group and the Radiation Protection Council has reviewed the proposed changes to the PULSTAR Technical Specifications as described in your letter of 30 April 1984. Based upon the recom-mendation of this review group, the Radiation Protection Council hereby approves these proposed changes for submission to the Nuclear Regulatory Commission. The Council does note, however, that modifications are to be made to the safety analysis portion of the proposed change to further support the proposed changes to Specification 3.6(a) and that these modifications are to be made prior to your submission of these documents to the Nuclear Regulatory Commission.
Thank you for your continued cooperation.
S ncerely, (fLg .'. h, t Donald E. Smith, Chairman Radiation Protection Council u .r. c... av,., % w u ,, . -o , .,, 1:. w ,i, ,, ,a .. .er .a o,.o,uw ,,, ,4 t I,a t I,,u er..r , < t \ ad. c r .t . .- I b -
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(viW m-un Dr. Donald E. Smith, Chairman Radiation Protection Council' 1607 Gardner Hall NCSU Campus
Subject:
Proposed Revisions to the PLA. STAR Technical Specifications.
Sir:
In accordance with Section 6.2.2 e. of the PULSTAR Technical Specifica-tions, it is requested that the Radiation Protection Council review and approve the enclosed revisions to the PULSTAR Technical Specifications.
This request supercedes and replaces any open correspondence to the Radia-tion Protection Council or the Nuclear Regulatory Commission (NRC) concerning proposed revisions to the PLA. STAR Technical Specifications.
Note that correspondence from Joab L. Thomas to the Nuclear Regulatory Com-mission dated July 10, 1979, and October 18, 1979, contained proposed revisions similar to those identified in this docenent. After literally years of waiting for the response from the WC, we were informed by the NRC that the request let-ter did not contain the appropriate nomenclature and, therefore, was set aside.
Specifically, the referenced correspondence did not explicitly say, " implement the following revisions", and consequently was deemed by the NRC to not be an official request.
Since this last submittal of proposed revisions to the Technical Specifica-tions, additional necessary changes have been identified, primarily due to organi-zational changes. It was deemed prudent to reinitiate the review and approval process for all of the proposed Technical Specification revisions due to the con-siderable amount of tims that has passed since initial conception of many of the proposed revisions. In addition, the NRC now requires a different format for the request that incorport 6's a page-for-page substitution rather than the previous 8
method of " Change this, to reads", etc.
Attachment I of this document includes the page-for-page revisions to the PULSTAR Technical Specifications. The revision bar in the right-hand margin in-dicates the particular section/ paragraph that is being changed. Note that in this proposed revision request, no pages were added or deleted from the present format of the PULSTAR Technical Specifications.
l North Carchroa $ tate linit,ornity le North Carolina's originallansi trant institution
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.: Dr.. Donald E. Smith, Chairman ~ Page.2 1 4
.Rediation Pyotection Council Attachment II of this document provides the Supporting Safety Analysis of .
the proposed revisions. Note that a reference to the page number and applicable l
. . section number-is provided to cross-reference the proposed revisions with the per- ,'
l tinent safety analysis.
Questions concerning the proposed revidions should be directed either to
- Mr. Thomas C. Bray, Reactor Operations Manager, or myself.
Your prompt attention to this-request would be greatly appreciated.
Respectfully submitted, d# f' Y Garry Dale Miller, L
Nuclear Operations Administrator, Associate Director GOM:1pe
Enclosures:
- 1) Attachment I - Revised Pages of the PLA. STAR Technical Specifications
- 2) Attachment II - Supporting Safety Analysis cc (all with attachments)
Dr. P. J. Turinsky Dr. J. Wortman, RSAG Mr. T. C. Bray Mr. T. L. Brackin L. Eudy, MP Adnin.
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'1.10 Measurina Channel - A measuring.channe' l -is the combination'of sensor, lines, amplifiers and output devices which are connected for.the pur-pose of measuring the value of a process variable.
~
.1.11~ Reactor Safety System - The reactor safety system is that specified combination of measuring channels and associated circuitry.which forms :
~the automatic protection systems'of the reactor, or provides informa . l tion which requires manus 1 protective action to be initiated.
1.12 Measured value - The measured value'of a process variable is the value of the. var: able as it appears on the output of a measuring channel.
1.13 True value - The true value of a process variable is its exact value at any instant. <
1.14 -Channel Check - A channel check is a qualitative verification of accept-able performance by observation of channel behavior. This verification l l shall include comparison of the channel with other independent channels i or methods of measuring the same process variable.
1.15 Channel Test - A channel test is the introduction of a known signal j into a channel to verify that it is operable.
1.16 Channe:, Calibration - A channel calibration 'is the adjustment of La i channe;,, such that its output responds with acceptable range and accuracy, to known values of the parameter that the channel measures.
Calibration shall encompass the entire channel, including equipment actuation, alarm and trips.
h 1.17 Reoortable Event - A Reportable Event is any of the following:
i a. Operation with any safety system setting less conservative than
- specified in the Limiting Safety System Settings section of the
, Technical Specifications; 1
- j. b. Operation in violation of Limiting Conditions for Operation i' c. Incidents or conditions which prevented or could have prevented
^
the intended safety function of an engineered safety feature or the reactor safety system; l d. Release of fission products from a fuel element '
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I 3.3 Reactor Instrumentation
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Acolicability These specifications apply.to the instrumentation which must be available and operable for safe operation of the reactor.
Ob_lective
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The objective is to require that sufficient 1nformation be available-to the operator to assure safe operation of the reactor.
Specifications The reactor shall not be operated unless the measuring channels listed in the following table are operable:
Minimum Measurina channel No. Operable
- a. Startup Power Level 1(a)
- b. Safety Power Level 1(D)
- c. Lineer Power Level 1(b)
- d. Pulse Energy 1(c)
- e. Primary Flow 1(d)
- f. FIow Monitor (flapper) 1
- g. Pool Water Temperature 1
- h. Pool Water Level 1 (a) Required only for reactor startup when power level
. is less than 4 watts. ,
(b) Bypassed for five (5) seconds during pulse and square wave . nodes.
(c) Required only in the pulse mode. Requires 1 out of 2 available channels (Pulse Energy or N-16 Channel).
(d) R ouired only for operation requiring forced convection cooling.
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.The neutron detectors and primary coolant flow meter provide assurance that measurements of the reactor power level, total energy and pri-mary coolant flow are adequately covered at low and high ranges. The '
reactor pool water level indicators provide early warning of the possibility of a leak =in the reactor cooling system or,the pool. The two channels available for determining total energy are: the N-16 integrator.
(total energy) and,-the pulse integrator (total energy) and pulse oscillo-
,-scope (peak power and total energyJ.- -
3.4 Reactor Safety Syst'em ,.
Apolicability-d These specifications apply to the reactor safety'systiem channels.
' Objective
- LThe objective is to require the minimum number of reactor safety system channels which must be operable in order to assure that the Safety Limits are not exceeded.
Soecifications .
The reactor shall not be operated unless the reactor safety system channels described in the following table are operable:
Minimum _ Operating Mode Measurina Channel No. Operable f;,nction in which Reo'd.*
- a. Startup Power Level 1 Inhibits Control Reactor Startup Rod withdrawal when neutron count is
< 2 cps.
- c. Linear Power Level 1(a) Scram All
- e. Pulse Energy 1(d) Provide total Pulse energy data on pulse (Manual <
Scram) l
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' Minimum Operating Mode in which req'd.
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Measurino Channel No. Doerable Function: .
. f. . Flow Monitoring 1 . Scram' All (Flapper)
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- g. - Primary Coolant Flow 1 ~ Scram. Steady State
> 250 kW3 all other modes
- h. Pool Water Temperature ~ 2 Alarm (Manual All
' Scram)-
- 1. -Pool Water Level 1 Scram All ,
J. Manual Button 1- Manual Scram All
- k. " Reactor on" Key- ~ 1 Manual Scram All switch
' 1. 'Over-the-Pool 1(a)(b) Alarm (Manual All Radiation Monitor Scram) .
(Bridge)-.
= m. . Bypass Timers 4(C) Limit bypass Pulse.and time on power- Square Wave level channels (a) Bypasses for five (5) seconds during pulse and square wave modes.
,' (b) Bypassed for less than one (1) minute during return of a pneumatic rabbit capsule from the core to the unloading station or three (3) during removal of experiments from the reactor pool.
(c) Activated during pulse and square wave modes, with two per applicable power level' measuring channel (safety power level and linear power level).
(d) In pulse mode only, requires either the Pulse Energy Channel or N-16 Channel to provide-information on pulse energy for manual scram.
Bases The inhibit function on the startup channel assures the required startup neutron source is sufficient and in its proper location for
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the reactor startup, -such that a minimum source multiplication count rate level'is being detected to Esure proper operation of the startup ,
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The reactor power level scrams provide the redundant protection
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U channels to assure that, if a condition should develop which would
.i tend-to cause the reactor to operate at an abnormally high power ,
level, an immediate automatic protective action will occur.to pre-vent the exceeding'of the Safety. Limit.
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The Pulse Energy Channel orIthe N-16 Channel provides the information
'on this parameter. In the event that an abnormal situation should develop, the operator is provided this information following the pulse-in order that he may return the. reactor.to its safest state, and take any other added precautionary action deemed necessary.
The alarms on the redundant pool water-temperature channels assure.
that the reactor operator has sufficient-time to take corrective action' if the temperature exceeds the;specified'11mit.
The primary coolant flow' scrams provide the redundant protection
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channels to assure when the reactor is at power levels which require forced flow cooling that,. If sufficient flow is not present, an ,
immediate automatic shutdown of the reactor will occur.to prevent the exceeding of a Safety Limit. The. Log N power channel is included in this section since it is one of the two channels which enables the two flow scrams when the reactor is above 250 kW.
The pool water level channel, along with the over-the-pool (Bridge)-
radiation monitor, provides two diverse channels for shutdown of the reactor and prevents the exceeding of the Safety Limit due to
- insufficient pool height (pressure).
To. prevent unnecessary. initiation of the evacuation confinement system during the return of the' pneumatic rabbit capsule from the core to the unloading station or during' removal of experiments from the reactor pool, the over-the-pool monitor may be bypassed during the specified time interval. A VAW radiation monitor that is located
- at the reactor pool bridge will continue to monitor radiation levels as a backup. This unit has also an audible alarm.
The manual scram button and the " reactor on" keyswitch ' provide _two .l manual' scram methods to the reactor operator if~ unsafe or abnormal conditions should occur.
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The redundancy of the bypass timers on each power measuring channel insures that no single time failure will negate the activation of the power level automatic protection systems.
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l 3.5 ~ RadiationMonitorinaEquipmenf.
, . Applicability
- This specification applies to the availability ~of radiation monitoring equipment which must be operable during reactor operation.
Objective To assure that radiation monitoring equipmentfis availablefor.evalu-ation of radiation conditions in restricted and unrestricted areas.
Specification The reactor shall not be operated unless the radiation _ monitoring equipment listed in the following table is operable.
- 1. Three fixed area monitors operating in the reactor building.(a)(b)(c)
- 2. Particulate andingas sampling air thebuilding facilityexhaust exhaustmonitors po,n)tinuously stack.tbhu (a) One of these monitors must be the over-the-pool (Bridge) radiation monitor. This monitor may be bypassed for ,
five (5) seconds during pulse and square wave modes.
(b) For' periods of time for maintenance to the' radiation monitoring channel, the intent of this specification will be satisfied if one of-the installed channels is replaced with a gamma-sensitive instrument which has its own alarm or which is kept under visual observation and if reactor operation is limited to the stcpify state mode.
(c) The over-the-pool monitor may be bypassed for less than one (1) minute during return of a pneumatic rabbit capsule i
from the core to the unloading station or three (3) minutes during removal of experiments from the reactor pool.
(d) May be bypassed for less than one (1) minute immediately after starting the pneumatic blower system.
Bases A continued evaluation of the radiation levels within the reactor building will be made to assure the safety of personnel. This is accomplished by the area monitoring system of the type described in Section 5.2.2 of the FSAR.-
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- A' continued evaluation of the' discharge air to.the environment will=
-be made using the information recorded from the particulate and gas
. monitors.
When the radiation ~ levels reach the alarm setpoint on any single area, " N or stack exhaust monitor,'the building will be automatically placed in confinement-as described in Section-5 of the FSAR.
. To prevent unnecessary initiation of the evacuation-confinement system-
- during the return of a pneumatic rabbit capsule from the core. to the-unloading station or during removal of experiments from the reactor pool, the over-the-pool monitor may be bypassed dJring the specified time. interval. A VAW radiation monitor that'is located at the reactor t
pool bridge will continue to monitor radiation levels as a back-up.
This unit has.also'an audible alarm.
- 3.6 ' Confinement .
Applicability ;
~ This specification applies to the operation of the reactor building confinement: system.-
Objective The objective is to assure that the confinement system is in operation-to mitigate the consequences of possible release of radioactive materials resulting from reactor operation.
I Specification
.. The reactor shall not'be operated unless the following equipment is
- operable,-and conditions met
- -
Operating Mode In Equipment / Condition Function Which Required
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- a. All doors, except the control To maintain reactor All room and basement corridor en- building negative l trance self-latching, self-- diffarential pressure.(f)
- closing; closed and locked.
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- b. Control room and basement - To maint_ain reactor (a) gil corridor entrance doors building negative ,
self-latching, self-- differential pressure.
closing and closed.
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-.;c. Reactor building under a: LTo maintain reactor (b)~ A11
. negative differential pres-- ; building negative '
1 sure of not less.than 0.2" . differential-pressure
.H20'with the normal venti-- with reference to out .
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lation system or 0.1" H 2O side ambient. :l with one confinement fen '
, operating.--
E d. Confinement initiation _ system Operable (c) All-2
- e. Evacuation system . Operable (d) - All-
- f. Magnahelic differential- ~ Operable- Pulse an'd square 1 vacuum gauges in control. wave room.
g.-Confinement filter trains. Operable (e) gli (a). Doors may be opened for personnel and equipment transport.
'between corridor area and reactor building.
(b) Except for periods of time not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to per-mit repair of system - during such repairs, the power will be limited to'100 kW and the steady state mode.
(c) Operability demonstrated also with auxiliary power source.
lt .(d) The public address system can serve temporarily for the reactor building evacuation system during short periods of maintenance.
(e) One filter train may be out of service ~for the purpose of main'-
1 tenance, repair, and/or surveillance for a period of time not to exceed 45 days. During the period of time in which one filter train is.out of service, the stan&y filter train shall be operating with the reactor building in the confinement mode.
(f). Doors may be opened by authorized persomel for less than (5) five minutes. for personnel and equipment transport provided audible and visual indication is available for the reactor operator..to verify door status.
~ Bases
'In the unlikely event of a fission product release, the. confine-
. ment l initiation system will secure the normal ventilation fans and close the normal inlet and exhaust dampers. In confinement, a N
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4.0_ SURVEILLANCE EQUIEENTS(l)
'4.1~ Fuel i
Applicability This specification applies to the survaillance requirement for the reactor fuel.
Objective The objective is to obtain data on length, diameter and bow of fuel pins for evaluation of long-term trends of fuel behavior.
Specifications a.. All fuel elements shall be visually inspected biennially but at intervals not to exceed nt - 1 (2&) months.
- b. At.least four fuel pins from two different pulsed fuel elements shall be inspected whenever a group of 100 pulses of greater than 1.0% Ak/k have been accumulated. One of these pins shall be from the highest flux region in the core.
Bases The biennial inspection of PULSTAR fuel pins has been shown to be adequate for Zr-2 cladded element's to insure fuel element integrity based on a long history of the prototype PULSTAR steady state and pulse operation.
The above inspection intervals are based on prototype experience.
The pins selected for inspcetion will be those most closely approsching the high energy density region of the core as determined from initial startup testing. As a minimum, the same pins will be measured at each inspection interval so that a trend can be established as a function of time. Baseline data for each fuel pin will be made prior to any usage in the reactor core.
- 1. All surveillance tests required by these specifications are scheduled as described; however, some tests may be postponed at the required intervals if that system or a closely associated system is undergoing maintenance.
In these cases, the surveillance test will be performed immediately after completion of maintenance and prior to reactor startup.
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- 4.2 TP ulse and Control Rods (Applicability
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This specification _ applies lto .the surveillance requirements for -
- the pulse and control rods.
Objective- '
The objective is to' assure the operability of.theLpulseLand control rods.
Specifications
- a. The reactivity worth of'the pulse rod and each con'rol t rod shall be determined Gnnually but at' intervals not to-exceed fifteen-(15) months for the steady. state' core in current use. The reactivity
-worth of the pulseLrod'and each control rod for the pulsing core
!' in' current use shall be determined within six (6) months prior to .
. pulsing operations.. The reactivity' worth of all rods shall be de-termined for any new-core or rod configuration, prior to routine operation.
- b. Control rod drop and drive times and the pulse rod drive time shall be determined annually but at intervals not to exceed fifteen-(15)- '
months, and after a control assembly is moved to a new position in
. the core or after. maintenance or modification is performed on the
> control rod mechanism. Pulse rod turnaround time shall be-deter-mined within six (6) months-prior to each pulsing operation.
- c. The pulse and control rods shall be visually inspected biennially l but at intervals not to exceed thirty (30) months.
Bases The reactivity worth of the pulse and control rods is measured to-assure that the' required shutdown margin is available, to provide a means for determining the reactivity worths of experiments inserted in the core, and to provide the pulse rod positioning for pulse and square wave operations. -The measurement of reactivity worths on an annual basis provides a correction for the slight variations expected due to-burnup. This frequency of measurement has.been found acceptable at similar research reactor facilities, particularly the prototype PULSTAR which has a similar slow change of rod value with burnup.
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1 : Control and pulse rod drive / drop / turnaround time measurements are made to determine whether the rods are functionally operable. Thesc
, ' time measurements may also'be utilized in reactor. transient _ analysis.
Control and pulse. rod visual inspections provide a method of detect-ing wear or corrosion in rod actuating mechanisms, and rod travel setpoints can'be verified during these-inspections.
Control and pulse rod surveillance procedure will document proper con-tro1~ rod system reassembly after maintenance and recorded post-main-tenance data will identify significant trends in rod performance.
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l 4.3! : Reactor Safety System
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. Applicability
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.* i cThis: specification applies'to the surveillance requirements for
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, :the Reactor Safety System.-
,Obiective
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The objective is to assure that the Reactor Safety System (RSS) will.
remain operable and will prevent the Safety Limits'from being' exceeded..
Specifications
- a. . A channel checkLof each measuring channel in the RSS shall be-
. performed daily when the reactor is in operation.
- b. A channel test of each channel in the RSS(a)'shall be performed '
prior to each day's ~ operation, or prior to each operation extending more than one day.
c.. A channel-calibration of the safety-and linear power level measuring channels by the calorimetric method shall be made.
semi-annually but at intervals not to exceed seven and one-half (7j) months.
~
,, d. A channel calibration of the following channels shall be made semi-annually but at intervals not to exceed seven and one-half (7j) lhonths.
- l. Pool Water Temperature '
- 2. Primary Cooling and Flow Monitoring (Flapper)
-3 . Pool Water Level
- e. A-channel calibration of the channel to be used for measuring pulse energy shall be made using a test pulse of.less~than 1%
ak/k reactivity insertion prior to any operation in the pulse mode with reactivity insertions above 1% ak/k. During this test pulse, the over-power trip bypass timers shall be verified to be operable.
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J (a) Testing the channel to be used.for measuring pulse energy-
.is required only prior to operation in the pulse or square.
. . wave mode. .
.t Bases ~
. The daily chame'l tests and checks will assure that the' safety; chamels are operable. The' calibrations at six (6) month intervals '
-will assure that~1ong-term drift of the channels is corrected. The
-calorimetric calibration of the reactor power level, in conjunction with the nitrogen 16 channel', will provide continual reference for'.
adjustment of the. linear, Log N and' safety channel detector positions
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and current alignment. _
4.4 Radiation Monitorina Equipment -
Applicability This' specification applies to-the surveillance requirements for-the area radiation monitoring equipment and the system for the stack radiation monitoring equipment.
-Objective The objective is to assure that the radiation monitoring equipment is operable.
. Specificiation' The area and stack' monitoring systems shall be calibrated annually but at intervals not to exceed fifteen (15) months. The'setpoints shall be verified weekly.
3 Bases _
These systems provide continuous radiation monitoring of the reactor building with a check of readings performed prior to and during reactor-operations. Therefore, the weekly verification of the setpoints in
- conjunction with the annual calibration is adequate to identify long-term variations in the system operating characteristics.
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l 4.5; Confinement System -
Applicability' This specification; applies to the surveillance.requiremen'ts for the confinement system.
Objective '
The objective is to assure that the confinement system is operable.
Specifications .
.a. The confinement and evacuation system shall be verified to be operable no more than seven (7) days prior to reactor operation,
- b. . Operation of the confinement system on auxiliary power will be checked every two weeks but at intervals not to exceed twenty-one (21) days.
~
- c. A visual inspection of.the door seals and closures, dampers and gaskets of the confinement and ventilation systems shall be per-formed semi-annually but at intervals not to exceed seven months to verify they are operable.
- d. The control room differential pressure gauges shall be calibrated annually but at intervals not to exceed fourteen months. -
- e. The confinement filter train shall be tested biennually but at intervals not to exceed twenty-six months; and prior to reactor operation following filter replacement.
- f. The 600 cfm air flow rate in the confinement stack exhaust duct shall be verified annually but at intervals not to exceed fourteen months.
Bases Surveillance of this equipment will-verify that the confinement of the reactor building is maintained, reference Section 5 of the FSAR.
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f6.0' ' ADMINISTRATIVE CONTROLS-
~'6.1 /Oraanization-6.1.1 The reactor facility shall'be an integral part of the Department of NL2 clear Engineering of'the School 1of Engineering of North Carolina State University.. The reactor shall be related to the.
University structure as shown in Figure 6.1-1.
~
.6.1.2 The Associate Director is responsible for the safe and efficient operation and. utilization of the PULSTAR reactor facility. -In_-
all matters pertaining to the operation of the facility and these Specifications, the Associate Director shall report.to and be directly responsible.to the Head, Department of Nuclear-Engineering. The minimum qualifications for the Associate Direc-tor are at least five years of reactor operating experience in-cluding at least two years of supervisory reactor' experience.
Beccalaureate or graduate study may be substituted ,for a maximum of one year reactor. operating experience.
6.1.3 The reactor shall'be under a Reactor Operations Manager who shall i be qualified as a licensed senior operator for this reactor. He shall be responsible for assuring that operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Radiation Protection Council.
6.1.4 There'shall be a Reactor Health Physicist responsible for assuring the safety of reactor operations from the standpoint of radiation protection. He shall function independently of the reactor opera-tions organization as shown in Figure 6.1-1. He shall be able to
. satisfy the qualifications of a Certified Health Physicist as set forth by the American Board of Health Physics or its acceptable equivalent. He shall be responsible for. reporting in accordance Section 6.7.5f.
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NUCLEAR REACTOR PROGRAM i"
ORGANIZATIONAL CHART-
'NCSU CHANCELLOR
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DEAN OF ,
ENGINEERING RADIATION !
PRDTECTION COUNCIL
~
NUCLEAR ENGINEERING DEPARTMENT HEAD RADIATION PROTECTION OFFICER I
DIRECTOR , NRP - 1 I-HEALTH l i' PHYSICIST I l ASSOCIATE l
- - L - .-- - .- DIRECTOR , NRP --------J 1
REACTOR OPERATIONS MGR.
- CHIEF CHIEF OF REACTOR REACTOR
, OPERATOR MAINTENANCE !
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.6.2_ -Review and Audit ~
w 6.2.1 There shall be a-Radiation Protection Council-(fPC)~whose-
-duties phall be to review and audit reactor operations, to ,
advise the Chancellor, North Carolina State University, and- .
to assure that'the facility is operated ~in a manner consistent
,with public safety'and within the terms of the. facility license.-
6.2.2 IFC' responsibilities shall includes
- a. Review proposed tests and untiried experiments which may.
. constitute an "unreviewed safety. question" pursuant to 107 R50.59. All:such reviews shall be accomplished with
- consideration of Sections 2 and 3 of these Technical-I Specifications.
- b. Review proposed changes to oceduhes,equipmentor. systems having safety significance ( or which may constitute an "unreviewed safety' question"; pursuant to 107R50.59.
i'
- c. Review facility Reportable Events defined in Section'1.17 of these Technical Specifications.
~
l 'd. Review proposed changes to the Technical Specifications or-.
Facility Operating License except for the PCSU PULSTAR REACTOR SECURITY PLAN which is withheld from public disclo-
'sure pursuant to 10 7 R2.790(d) and 107 R73.21.
- e. Review and audit facility' operation, including equipment-
! performance, operating personnel, operating records and results of surveillance tests and inspections.
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.l. -Equipment or systems having safety significance are defined as those L channels and systems identified in Section 3, Limiting Conditions of l Operations, in particular 3.3, 3.4, 3.5, and 3.6 L
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, 6.2.3. The Radiation Protection Council shall have at least five members, of whom no more than the.. minority shall be from the line organization shown in Figure 6.1-1. In addition to the j Radiation Protection Council,,three personc are appointed -
by the _ Chancellor, tpon the recommendation-of the PC, to '
form a Reactor Safeguards Advisory Group (RSAG). This group serves as a permanent committee to the P C and will-be solely responsible for independent appraisals of reactor operations, and reporting the results of its investigations to the PC, Department Head and Associate Director. The PC and RSAG shall be made up of senior faculty and staff who shall collec-tively provide experience in reactor engineering, reactor operations, chemistry and radiochemistry, instrumentation and control systems, radiological safety and mechanical and electrical systems. The campus Radiological Safety Officer (RS0) shall be a permanent member of the P C.
s
. 6.2.4 The Radiation Protection Council shall have written statements on its responsibilities and authority.
6.2.5 Minutes of each PC meeting shall be distributed to the Chancellor and all PC members. Minutes of each RSAG meeting
.shall be distributed to the PC; Head, Department of Nuclear Engineering and Associate Director. l 6.2.6 A quorum shall consist of not less than a majority of the full RPC or RSAG, and shall include the chairman or his designated alternate. Only a minority or less of the line organization shown in Figure 6.1-1 shall be present in any quorum.
6.2.7 The P C shall meet at least every four calendar months, and upon call of the Chairman; while the RSAG shall meet at least every six calendar months and upon call of the Chairman.
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- 6.3> Operatina' Proc'edures-La. ;0perating procedures shall be written, todated periodically and Lfollowed for.these operations
~
e s 1. Normal startup, operation', and shutdown of the reactor and components _ involving nuclear safety of the systems
- 2. Refueling' operations and installation and removal of Control
- ' Rods'and experimental facilities .
3.- Actions to' be taken to' correct specific and foreseen potential-
. malfunctions of systems or cuTpenents, includ_ing responses to
. alarms, suspected primary coolant system leaks,.and abnormal reactivity changess ,
- 4. Emergency conditions involving potential or actual release
.L of radioactivity, including provisions-for evacuation,~
l re-entry, recovery, medical supports L -5. Preventive and corrective maintenance operations which
- could have an effect on reactor safety 4
- 6. Civil disturbance on or near the. campus:
- 7. Periodic surveillance including channel test and channel
- calibration of reactor instrumentation and Reactor Safety Systems, area monitors and continuous air monitors.
+
- 8. Radiation control.' These procedures shall be maintained and be available to all operations personnel.
- 9. Review and approval of changes to operating proceduress a
- 10. Preparation, approval and periodic review of facility Emergency
, Plan and Security Plan.
4
- b. Substantive changes to the above procedures shall_be made only with j .the approval of the Radiation Protection Council. Temporary changes
.to the procedures that do not change their original intent may be
~made with the approval of the Associate Director. All such temporary
- changes to the procedures shall be documented and subsequently re-viewed by the IPCs
- c. Drills on emergency procedures shall be conducted annually.
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l 6.4 Action to be Taken in the Event a Safety Limit is-Exceeded l
In the event-a Safety Limit.is exceeded, or. thought to be exceeded:
- a. The reactor;shall be shut down and reactor operations shall ' )
.not be resumed until authorized by the Nuclear Regulatory. 1
.Connission;
'b. An immediate report of the' occurrences shall be made to the Chairman of the Radiation Protection Council, and reports shall be made to the Nuclear Regulatory Commission in accor-dance with Section 6.7 of these specifications; and
- c. A report shall be made which shall include an analysis of the causes and extent of possible damage, efficacy of cor-rective action, and recommendations for measures to prevent the probability of recurrence. This report shall be sub-mitted to the Radiation Protection Council for review, and a suitable.similar report submitted to the Nuclear Regulatory Commission when authorization to resume operation of the reactor'is sought.
6.5 Action to be Taken for Reportable Events
~
In case of a reportable event, as defined in Section 1.17 of these specifications, the following action shall be taken:
- a. The' Reactor Operations Manager, Associate Director, Reactor Health Physicist and Radiation Protection Council shall be notified of the occurrence. Prior to resumption of operations, action shall be taken to correct the initiating' condition and to prevent its recurrence.
- b. A report shall be made to the Radiation Protection Council for review. The report shall include an analysis of the cause of the occurrence, the ~ effectiveness of corrective action taken, e and recommendations for measures to prevent or reduce the pro-bability of recurrence.
- c. A report shall be submitted to the Nuclear Regulatory Commission in accordance with Section 6.7 of these specifications.
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J6.6 Plant Operatina Recor'ds' ;
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-In addition.to the requirements'of applihable regulations and in no -l
.way-substituting therefor, records and logs of the following items, as a minimum,' shal1~be kept in a manner convenient for review and ,
shall be retained as indicated:
Records to be retained for a period of at least five (5) years:
a.'
'1. . Normal plant operation.and maintenances
- 2. Principal maintenance activities
- 3. Reportable eventsp.
, ;4. ; Equipment and components surveillance activities;
- 5. Experiments performed with_the react'org
- b. ' Records to be retained for the life of the facility:
- 1. Gaseous and 11guld radioactive waste released to the environs;
- 2. Off-site environmental monitoring surveys
- 3. Radiation exposures for all PULSTAR personnel;
- 4. Facility radiation and contamination surveys;
- 5. ' Fuel inventories and transfers:
- 6. Updated, corrected, and as-built facility drawings.
6.7 Reportina Re.Quirements In addition to the requirements of applicable regulations and in no
. way substituting therefor, reports shall be made to the tRC as follows:
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- 6.7.'l Within 2'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,~ a telephone'or telegraph report to the.
g' 4 USE. Region II Regional Adninistrator of:
.a. ~Any accidental off-site release of radioact ii v ty.above permissible _ limits, whether or not1the release resulted # '.
~1n property. damage,' personal injury or exposure;
- b. 'Any 'significant variation of measured values from a cor-
. responding predicted or previously measured value of-safety-related operating parameters occurring during operation of the reactor.
- c. Any Reportable Event as defined in Section 1.17 of these specifications; and
- d .~ -Any violation of a Safety. Limit.
6.7.2 Within 10 days, a written report to.the USMIC Region II Regional Adninistrator and a written. report ccpy to the Director of the of the USMtC Office of Nuclear Reactor Regulation.of:
- a. Any significant variation of measured values from a cor-responding predicted value of previously measured value of. safety-related operating characteristics. occurring
, during operation of the reactor;
- b. Incidents or conditions relating to operation of the t
facility which prevented or could have prevented the performance of engineered safety features as described in these specifications;
- c. Any Reportable Event as defined in Section 1.17 of .
these specifications; and
- d. Any violation of a Safety Limit. 1 6.7.3 Within 30 days, a written report to the USNRC Region II Regional Adninistrator and a written report copy to the Director of the f USMtC Office of Nuclear Reactor Regulation of I
- a. Any substantial variance from performance specifications ;
contained in these specifications or in the Safety Analysis '
Report; l 1 1 o
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.b.;:Any significant change in the transient or accident-
- analyses as' described'in the~ Safety. Analysis' Report;
- c.- Any changes in facility. organizations.and
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- d. . Any observed inadequacies in the implementation of.
administrative or procedural' controls. -
6.7.4- A report within 120' days until October.30,'1973, then 60 days thereafter following completion of startup testing .
of the reactor (in writing to the Director'of the USteC Office
-of POclear: Reactor. Regulation) upon receipt of a new facility license, or an amendment to the license. authorizing an increase in reactor power level, describing the measured values of the-operating conditions or characteristics of the reactor under under the new conditions including:.
- a. An evaluation of the safety analysis ~ submitted with the license application in light of measured operating charac-teristics when such measurements indicate that there may i' be substantial variance from analysis.
b
- b. A reassessment of the safety analysis submitted with the license application in light of measured operating charac-
^
teristics when such measurements indicate that there may be substantial' variance from prior analysis.
6.7.5 An annual report within 60 days'following the 30th of June of each year (in writing to the Director of the UStRC Office.of Nuclear Reactor Regulatiorp providing the following information:
with wPy to tac RepE %gia wl thMs1Mor
, a. A brief narrative summary of (1) operating experience in-
- cluding a cross section of experiments performed, (2) i changes in performance characteristics related to reactor 2
safety and occurring dJring the reporting period, and (3) results of surveillance tests and inspections;
- b. Tabulation of the energy output (in megawatt days) of the reactor, amount of pulse operation, hours reactor was r
critical, and the cumulative total energy output since initial criticality; l
- c. The number of emergency shutdowns and inadvertent scrams, including reasons therefor; i
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(a) Method of disposal.-
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.(b)1 Total radioactivity I'n the tank (in microcuries) prior
- to disposal.-
- . (c) Total volume of liquid in tank (in liters). '
- (d) . The dried residue of ~a' one (1) liter sample shall be analyzed for the principal gamma-emitting radionuclides.
.The identified. isotope composition with estimated concentrations shall be reported. The tritium content ;
shall be included.
Gaseous Waste (sumarized on'a monthly basis)-
- 1. Radioactivity discharged during the reporting period (in.
curies) for (a) Gases.
(b) Particulatesi with half lives greater than eight days.
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The W C used and the cstimated activity (in curies)
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' discharged during the reporting period, by nuclide,
'j for all gases.and particulates based on representative -
isotopic analysis. -l i.
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-. Solid Waste
- 1. The total amount of solid waste packaged (in cubic '
feet). ,
- 2. The total activity involved (in curies).
- 3. Theidates of stilpment and disposition (if shipped off-site).
'g. A summary of radiation exposures received by facility' '
personnel and visitors, including pertinent details of significant exposures.
- h. A summary of the results of radiation and contamination
- surveys performed within the facilityg and ' ;
- 1. ~A description of any environmental surveys performed outside the facility.
1- 6.8 Review of Experiments ,
6.8.1 All proposed experiments utilizing the PULSTAR reactor shall i be evaluated in writing by the experimenter and by a licensed l Senior Reactor Operator staff member approved by the Associate Director. The evaluation of the experiment shall be reviewed i by the Reactor Operations Manager (and by the Reactor Health Physicist when appropriate) to assure compliance with the pro-4 visions of the facility license, these Technical Specifications, and the M E regulations. If the Reactor Operations Manager and the Reactor Health Physicist determine that it is a tried experi-i ment, it may be approved by the Reactor Operations Manager or Associate Director. The evaluation of the experiment shall include:
, a. The reactivity worth of the experiments
- b. The integrity of the experiment,' including the effects of i changes in temperature, pressure, chemical composition,
, or radiolytic decomposition.
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'c. . Any.physicaliorLchemical interaction'which could occur 1 with the. reactor components and~
.d. Any. radiation hazard that may' result from the activation of materials or:from external beams. '
~
!If it-is the opinion of any'one of the above', that the experiment does not meet the criteria, it shall be submitted,1 by the experimenter, to the Radiation Protection Council as.
an " untried experiment" for approval and issuance of a ~
- project. number.
6.8.21 Prior to performing an untried experiment-in the reactor,.
It shall be reviewed and approved:in writing by the IPC,
, following their determination that it does not' involve'an' unrevieweo safety question. The IFC review shall consider-the following:
~
y i a. The' purpose of the experiment;'
- b. A procedure for the performance of the experiments
- c. Evaluation made as in paragraph 6.8.1 abover and
- d. Evaluation approved by' the Associate Director.
- 6.8.3 In evaluating experiments, the following assumptions shall be-used for the purpose of determining that failure of the experi-
! ment would not cause the appropriate limits of.-10 T R 20 to be exceeded:
^
- a. If the possibility exists that airborne concentration of radioactive gases or aerosols may be released within the confinement, 100% of the gases or aerosols will escapeg
- b. If the effluent exhausts throu@ a filter installation designed for greater than 995 efficiency for 0.3 micron
- particles, at least 105 of the particulates will escapes and
- c. For a material whose boiling point'is above 1300F and where vapors formed by boiling this material could escape only through a volume of water above the core, at least 10% of j' these vapors will escape.
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4 yb U. Docket facility'No'.nR-120 Noe 3P37
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Pt1 STAR TECHNICAL: SPECIFICATION EVISION
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SLFPORTI M SAFETY ANALYSIS ; '
^
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'ITEMD 'PAGE# -SECTION- SAFETY ~ ANALYSIS
.. (1)~ 22 - 1.17- - Thelterm " Abnormal Occurrence": has been i-I
- replaced'by " Reportable Event":per the Estandard use by the nuclear industry.-
-- Note that this change is an editorial-change'and does not affect reactor safety.
(2) 2- 1.17 c.' __ . Purpose of rewriting Para. 1.17-c is to make .
make-statement conform more nearly to that of
, ANSI /ANS-15.1-1982._. Sense of the statement-is unchanged but language and terminology has been improved.
(3)- 17 - 3.3 d. .The N-16 Channel Integrator reliability has been demonstrated in our Start-Lp program and subsequent routine pulsing. The N-16 Channel respords in the same manner as the :
Pulse Energy _. Channel, providing a direct measurement of pulse energy;following the-pulse. The requested changes'for Sections
- 3.3 and 3.4 identify the N-16 Chamel as an available channel for measuring pulse energy.
The PULSTAR Final Safety Analysis Report ap-proves the use of the N-16 Chamel for Pulse energy measurement in Section 7.1.4. . Note that the Commission had previously approved the requirement of_only one pulse energy.
chamel by virtue of' footnote (c) to Section 3.3; therefore, the requested change is edi-torial in nature.
(4)' 17- 3.4 foot- Additional clarification has been added of note (c) the available pulse energy chamels.- Note that the analysis of item (3) above applies.
(5)' 18 3.4 e The specified change adds.a footnote clari-fying the choice of pulse energy chamels available cbring pulsing operations. Note that the analysis of item (3) above applies.
(6) 19 3.4 foot- Same as for item (4) above.
note'(d) -
(7) !20 3.4 Bases Same as for item (3) above.
3rd Paragraph)
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- Facility.No, R-120 F 'N! '
ya Docket Noi- '50-297- C'
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a PtLSTAR TECHNICAL SPECIFICATION REVISION -
SLPPORTIfC SAFETY ANALYSIS ,
ITEM
- PAGE* SECTION -
... SAFETY ANALYSIS
=(8)- 21 3.5.2 :
A . footnote (d) has been added allowing the - 0 4
- bypass 'of the specified monitors cbring the initiation of the pneumatic blower system.
The requested change will prevent the un- <
necessary alarm which initiates the evacua--
tion and confinement system. .The initiation of the. pneumatic blo'wer. system during power i operation results in a small volume of air
. with an Argon-41 activity. level, above the normal setpoints, moving through the ventila-tion system causing the stack exhaust monitors to alarm.. The requested change will-inhibit the unnecessary evacuation during the short-term transient of stack exhaust activity without compromising reactor safety. In the event that a fission product release should occur simultaneously with the bypass of the specified monitors, the over-the-pool monitor would still be capable of initiating evacua-l tion and confinement. The bypass would.be performed by momentarily depressing the Fall /
Reset button on the specified monitors. 'l$on
- release of the Fail /Meset button, the channel automatically resets itself to the alarm-ready status. This bypass action only bypasses the alarm initiation and not the monitoring func-
- tion of the channels. Specifically, the three
' channels normally used for monitoring stack ;
exhaust activity, i.e. Stack Gas, Particulate '
and the Auxiliary GM, are all recorded for sub-sequent calculation of stack exhaust releases.
i (9) 21 3.5.2(c) Changed word "is" in first line to "may be".
, Original language implied an automatic func-tion. Function has always been manual, at the 1
- discretion of the reactor operator.
l (10) 22 3.5 foot-. Same as for item (8) above.
j note (d) l l
- (11) 22 3.5 Bases Added word " alarm" to describe type of set-point that, when tripped, will autortatically {
l 1 7 place the reactor building in confinement.
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' /' Facility'No. R-120 . ~
- gf: Docket,No.. 50-297- .
y~ _ ~c
'" PtLSTAR TECHNICAL SPECIFICATION EVISION StPPORTING SAFETY ANALYSIS e
- f
. ITDW PAGE* SECTION SAFETY ANALYSIS
- - L(11)- 22. .3.5 Bases- Changed language also to specify that Jan (continued) sinole area or stack exhaust monitor wi n pro-vide the automatic action. 4 (12) 23 3.6 a. A footnote that allows the specified doors to be opened with certain restrictions has been added.
- Maintaining the reactor building differential pressure while operating is the responsibility of the reactor operator. The requested change will allow any reactor bay door to be opened for periods of time to permit transport of
-equipment and personnel. In the event a radio-active release should occur, the evacuation and confinement system would be placed in operation at the direction of the reactor operator and if the reactor bay door had been open for person-
, nel or equipment transport, the door would self-close upon exit of personnel, thereby maintaining required negative differential pressure in the confinement mode. Furthermore, audible and visual indication of the door status will enable 4
the operator to verify that the doors are closed in accordance with operating procedures follow-inij the initiation of the evacuation and confine-
,. ment system.
(13) 24 3.6 foot- Same as for item (12) above.
note (f)
(14) 28 A.2 a. It is North Carolina State's intention to adapt its Technical Specifications to ANS-15.1A078-1974, " Standard for the Development of Technical Specifications for Research Reactors," and toward this end, a tolerance interval has been added to the surveillance intervals. This tolerance shall provide for continuity of surveillance test, main-tenance, and reactor operations. Based upon re-sults of proven satisfactory performance of the current PtLSTAR Surveillance Program, adding a tolerance interval to the subject specifications will not compromise reactor safety. The request-ed tolerance intervals are consistent with Section 4 of ANS-15.1A078-1974, e.g. an annual surveil-p .
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" + Facility! No.~ R-120]@87 Docket;No.
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.'PULSTAR TECHNICAL SPECIFICATION .EVISION .
SLPPORTING SAFETY ANALYSIS ITEM #' PAGE4> -SECTION- SAFETY ANALYSIS-
- (14).
' ' 28c 4.2 a. lance item has two-month tolerance, etc.-
(continued) .
L(15) ~ 29 4.2. a.- The terms " standard. reference core"1and ' pulse
~ core" have been replaced by'" steady state core Lin ~ current use" and " pulsing core in current-use", respectively, in order to adag: 'ha sur .
veillance requirement to the latest core confi-
'~
guration-in operation. The rod worth measure-ments for.the pulsing core within six months prior..o the pulsing insures that recent data will be available to precisely position the pulse rod, and the surveillance'of the steady state core.
will provide adequate Information about. changes.in rod worths due to burnup.
(16) 29 a.2 b. 'The requirement for: measuring pulse rod drop-time has been removed since this measurement has no physical significance. Proper movement of the pulse rod'is. tested by pulse rod turn-around time measurements. The addition'of a tolerance on the surveillance interval is con--
[ sistent with the safety analysis of-item (14) above.
(17) 29 A.2 c. Safety analysis consistent with item (14) above.
(18) 30 4.2 Bases Safety analysis consistent with item (16) above.
(19) 30 4.3 c. Safety analysis consistent with item -(14)~ above.
(20) 30 4.3 d. Safety analysis consistent with item (14) above.
(21) 30 4.3 e. The requested change for Section 4.3 e. Insures that the appropriate surveillance test shall be performed on the channel that is to be used for measuring W u e.nergy (either the Pulse Energy Channel M k d Channel). Analysis for use of-the . V hs , al to measure pulse energy is con-sist.w .h ne safety analysis of Section 3.4.
-(22) 31 4.3 foot- Safety analysis consistent with item (21) abcfve.
note (a)
(23) 31 4.4 Safety analysis consistent with item (14) above.
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..iFabilitycNo.R-120'
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I Docket.No.; ~ 3@87 , _ n [m ,
- F iPlLSTAR'TEO94ICAL SPECIFICATION FEVISION >)
SLPPORTIls' SAFETY ANALYSIS ITEM 8' iPAGE+ -SECTION- SAFETY ANALYSIS (24)~ 132 :4.5 a. The proposed change provides for a more
' realistic test of-the confinement'initia-~
tion and evacuation system ~on the auxiliary
. generator. The present specification for-
. testing the confinement initiation system on the auxiliary generator requires a line-up of circuit breakers that would normally not be present during reactor operations.
This is due to the fact that upon loss of commercial power to the Control Room Dis-tribution Panel,"the confinement ~ initiation relay de-energizes so that upon, return of corimercial power or auxiliary generator power, the reactor building is automatically placed
, in the confinement mode. Therefore, the re-quested ch.rs.5 to Sections 4.5 a. and 4.5 b.
remove the requirements to test-the confinement initiation system on the auxiliary generator; The evacuation system, including logic relays and evacuation' horns, is powered from the Con-trol Room Distribution Panel. There is no .
dedicated bus.from the generator to the evacu-ation/ confinement initiation system,:therefore, if the confinemeric system will initiate on com-mercial power, then it will also initiate on auxiliary power. Consequently, the ability of the evacuation / confinement initiation system to operate on the auxiliary generator is assured by the combination of two tests: 1) the test de-monstrating that the Control Room Distribution Panel can be powered from the auxiliary generator and 2) the weekly test ~of the evacuation / con-finement initiation system on the console power via commercial powcr to the Control Room Distri-t bution Panel.
( 25). 32 4.5 c. This surveillance requirement has been found more convenient to perform on a semi-annuel basis. Since the requested change reduces the-surveillance intervals, reactor safety is not compromised.
4 (26) 32 4.5 d. Safety analysis consistent with item (14) above.
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, Occket.No. 15~3 7 PtLSTAR TECHNICAL SPECIFICATION REVISION =
SLPPORTING SAFETY ANALYSIS- 4 l
I19 # PAGEC .SECTION SAFETY' ANALYSIS (27) .32 4.5 e.: ' Safety. analysis consistent with-item (14) above.
-(28) 32 " 4.5 f. Safety analysis consistent with ites (14) above.
(29) 36 .6.1.2~ A re-organization of the f4Jclear Reactor Pro- )
gram has resulted in the change of titles for the Nuclear Operations Adninistrator and Reactor Supervisor positioris.
. Responsibilities / qualifications for these posi- I tions have remained unchanged; therefore, this change is only editorial in nature. Note that l positions in the PAJclear Reactor Program that 1 are not directly related to the PtLSTAR Reactor H Operation have been intentionally omitted from !
. Section 6.0 of the Technical Specifications.
1(30) 36 6.1.3 - Consistent with the safety analysis of item (29) above.
(31) 36 6.1.4 - As part of the re-organization described in item (29) above, the " Technical Services Group" has been integrated into other divisions of the Nuclear Reactor Program. The Reactor Health Phy-sicist position, however, remains as an inde-pendent review of the reactor operations. The 4
Reactor Health Ph sicist reports to the Head, Department of Nuclear Engineering.
(32) 37 Figure . Safety analysis consistent with items (27) and 6.1-1 (29) above.
(33) 38 6.2.2 The intent of rewriting 9 6.2.2 under Review and Audit is twofold: To adhere more closely to the guidelines of ANS-15.1/ ANSI N378-1974 and to make editorial changes for the purpose of clarity.
All the review and audit features of the original Technical Specifications have been retained while the actual number of separate statements under Para. 6.2.2 have been reduced. All references to
" abnormal" occurrences, conditions and operations have been re-edited to categorize these instances as " reportable events", in keeping with ANS 15.1
- terminology.
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. PULSTAR TEClflICAL SPECIFICATION FEVISION' SLPPORTING SAFETY ANALYSIS;'a
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- PAGE* .SECTION SAFETY ANALYSIS:
(34) 39 6.2.3, Substituting word " operations" for " procedures" i line 8' gives RSAG broader review authority. Reactor Operations' includes procedures and other audit areas as well.. Word change-conforms to termi-nology identified in ANS 15.1.
(35) 39 '
6.2.3, Safety analysis consistent with' item (29) above. -
line 10' (36)- 39 6.2.5 Safety analysis consistent with item-(29) above.
(37) 40 6.3 b. Safety. analysis consistent with item (29) above.
(38)-. 41 6.5 a. Safety analysis consistent.with item (29) above.
- (39) .,
41 6.5 c. Change is editorial in nature only.
(40). 42 6.6.a.3. Safety analysis consistent with item (1) above.-
(41) 42 6.7 Change is editorial in nature only.
(42) 43 6.7.1 c. Safety analysis consistent with item (1) above.
(43) 43 6.7.2 c. Safety analysis consistent with item (1) above.
(44) 44 6.7.5 a. The requested change clarifies the necessary reporting _ requirements by removing the duplica-ti m between 6.7.5a and 6.7.5e. In particular, changes in facility design and changes in opera-ting procedures were previously included in both l
sections. Reporting requirements for experiments have also been clarified, i.e., a cross-section of experiments performed is now reported for Sec-tion 6.7.5a and the safety. evaluation and descrip-tion of new experiments and tests are reported in Section 6.7.5e.
(45) 45a Continuation The letter g. was removed from the " Gaseous Waste L
of 6.7.5.f (Summarized on a Monthly Basis)" as an editorial correction. Gassous waste falls under the general category of radioactive effluents detailed in 6.7.5 f.
(46) 46 6.7.5 g,h,1 The requested changes to 6.7.5 g, 6.7.5 h, and the N
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Docket No.. 3D37 .,
PULSTAR TECHNICAL' SPECIFICATION FEVISION .
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StPPORTING SAFETY ANALYSIS
' ITEMD ' PAGE# 'SECTION SAFETY' ANALYSIS (46)~ .46 6.7.5 g,h,1- addition; of 6.7.51. clarifying the reporting
-(continued) requirements for surveys and exposures. The requested chang to 6.7.5 g. replaces the words "date and time with " pertinent details". This
, change will provide for an increase in rcporting requirements from just date and time to more sig-nificant details such as when, where and how the exposure occurred. .The latter details are more
-- important in terms of reporting a significant exposure rather.than just when it occurred.
(47) 46 6.8.1 Safety analysis consistent with item (29) above.
(48) 47 6.8.2 Safety analysis consistent with' item (29) above.
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