ML20005D915

Amend 11 to Tech Specs for Pulstar Reactor Re Relicensing

ML20005D915
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Site:	North Carolina State University
Issue date:	12/15/1989
From:	North Carolina State University, RALEIGH, NC
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NUDOCS 9001020247
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'l APPENDIX A FACILITY LICENSE NO. R-120 TECHNICAL SPECIFICATIONS FOR THE.

L" NORTH CAROLINA STATE UNIVERSIIY

[

PULSTAR REACTOR l-DOCKET NO. 50-297 ORIGINAL ISSUE DATE: August 25, 1972 l'

AMENDMENT 11 - December 15, 1989 L

l l

l 9001020247 891218 PDR ADOCK 05000297 P

PDC

Appendis A Technical Specifications TABLE OF CONTENTS hEt 1.0 DEFINITIONS.........................................

1 2.0.

SAFETY LIMITS AND UMITING SAFETY SYSTEM SET 1'INGS

... 5 2.1 Safe ty Limits..................................... 5 2.1.1 Safety Umit for Forced Convection Flow............... 5 2.1.2 Safety Umit for Natural Convection Flow.............. 7 2.2 Umiting Safety System Settings......................... 8 2.2.1 LSSS for Forced Convection Flow.................... 8 2.2.2 1.SSS for Natural Convection Flow.,................. 9 3.0 LIMITING CONDITIONS FOR OPERATION..................

10 3.1 Reactor Coolant Inlet Temperature......................

10 3.2 Reactivity...................... ;................

11 33 Reactor Safety System..............................

13 3.4 Reactor Instrumentation.............................

15 3.5 Radiation Monitoring Equipment............,,..........

16 3.6 Confinement and Main HVAC Systems...................

18 L

3.7 Limitations on Experiments........................... 20 3.8 Operation with Fueled Experiments...................... 23

[

3.9 Primary Coolant

...............................25 4.0 SURVEILLANCE REQUIREMENTS......................... 26 I

4.1 Fu el......................................... 2 6 4.2 Control Rods.................................... 27 43 Reactor Instrumentation and Safety Systems................. 29 L

4.4 Radiation Monitoring Equipment........................ 31 4.5 Confinement and Main HVAC Systems.................... 32 4.6 Primary and Secondary Coolant........................33-5.0 DESIGN FEATURES................................... 33 5.1-Reactor Fuel

....................................34 5.2 Reactor Building.................................. 34 53 Fuel Storage

....................................34 l.

5.4 Reactivity Control.................................. 34 5.5 Primary Coolant

..................................35 L

6.0 ADMINISTRATIVE CONTROLS........................... 36 6.1 Organization..................................... 3 6 l

6.1.1 Organizational Structure......................... 36 6.1.2 Minimum - Staffing............................. 37 6.13 Class A Reactor Operator Duties................... 38 6.2 Review and Audit................................... 39 l

6.2.1 Radiation Protection Council

.....................39 6.2.2 RPC Composition and Qualifications................. 39 6.23 RPC/RSAG Review Function...................... 39 6.23 RSAG Audit Function

.........................40 63 Operating Procedures............................... 42 6.4 Experime nts...................................... 4 2 6.4.1 New or Untried Experiments

....................43 6.4.2 Tried Experiments............................. 43 6.5 Action to be Taken for a Safety Limit Violation............. 44 Amendment 11 i

December 15,1989

't Appendix A Technical Speelnentions TABLE OF CONTENTS (continued) 6.6 Action to be Taken for a Reportable Event (other than a SL Violation)..................................... 4 5 6.7 Reporting Requirements............................. 46 6.7.1 Reportable Events............................ 46 6.7.2 Permanent Changes in Facility Organization............ 46 6.7.3 Changes Associated with the Safety Analysis Report....... 46 6.7.4 Annual Operating Report......................... 46 6.8 Retention of Records

..............................49 2.1-1 Power-Flow Safety Limit Curve 6.1-1 NCSU PULSTAR Reactor Organization l..

I l

Amendment 11 il December 15,1989

.. =

Anpendis A Technical Speelfications included in this document are the Technical Specifcations and the " Bases"for the Technical Specifcations. These bases, which provide the technical suppon for the individual technical specifcations, are included for information purposes only. They are.

not part of the Technical Specifcations, and they do not constitute limitations or requirements to which the licensee must adhere.

1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Condition of Operation (LCO) are as defined in Paragraph 50.36 of 10 CFR Part 50.

1.1 Class A Reactor Onerator: A Class A reactor operator is an individual who is certified to direct the activities of Class B reactor operators. Such an individual is also a reactor operator and is commonly referred to as a Senior Reactor Operator (SRO).

1.2 Class B Reactor Onerator: A Class B reactor operator is an individual who is certified to manipulate the controls of a reactor. Such an individual is commonly referred to as a Reactor Operator (RO).

1.3 Channel

A channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.

1.4 Channel Calibration: A channel. calibration is an adjustment of a channel, such that its output responds with acceptable accuracy to known values of the parameter that the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trips and shall be deemed to include a Channel Test.

1.5 Channel Check: A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems of measuring the same variable.

1.6 Channel Test: A channel test is the introduction of a known signal into a channel to verify that it is operable.

1.7 Cold Critical: The condition of the reactor when it is critical, with negligible xenon, and the fuel and bulk water are both at an isothermal temperature at 70*F.

1.8 Confinement

Confinement means a closure on the overall facility which

]

controls the movement of air into it and out through a controlled path.

1.9 Control Rod: A control rod is a' neutron absorbing blade having an in-line drive which is magnetically coupled and has SCRAM capability.

Amendment tt 1

December 15,1989

' Appendis A Technical Specincations 1

1.10 Excess Reactivity: Excess reactivity is that amount of reactivity that would exist

)

if all Control Rods (and Shim rod) were fully withdrawn from the point where the reactor is critical (k,=1).

1.11 Experiment: Any operation, hardware, or target (excluding devices such as detectors) which is designed to investigate reactor characteristics or which is intended for irradiation within the pool, on or in a beam tube or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design. Specific categories of experiments include a.

Tried Experiments: Tried experiments are those which have been previously performed in this reactor. Specifically, a tried experiment has a similar size, shape, composition and location of an experiment previously approved and performed in the reactor, b.

Secured Experiment: - A secured experiment is any experiment, L

experimental facility, or component of an experiment that is held in a L

stationary position relative to the reactor by mechanical means. The L

restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.

c.

Non. Secured Erneriment: A non-secured experiment is an experiment that does not meet the criteria for being a " Secured" experiment.

I d.

Movable Erneriment: A n$ovable experiment is one where it is intended l

that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

e.

Fueled Erneriment: A fueled experiment is an experiment which contains fissionable material.

1.12 Ernerimental Facilities: Experimental facilities are facilities used to perform experiments. They include beam tubes, thermal columns, void tanks, pneumatic transfer systems, in core facilities at single assembly positions, out of-core irradiation facilities, and the bulk irradiation facility.

1.13 Measured Value: The measured value is the value of a parameter as it appears on the output of a channel.

l 1.14 Operable: Operable means a component or system is capable of performing its l

intended function.

1' l

1.15 Operating: Operating means a component is performing its intended function.

1.16 Oneration: Operation is any condition when the reactor is not secured.

Amendment 11 2

December 15, 1989

~__ _.

Appendix A Technical Specifications 1.17 acm: A unit of reactivity that is the abbreviation for " percent milli" and is equal to 10' Ak/k reactivity. For example,1000 pcm equals 1.0% ak/k.

1.18 Reactor Onerator Assistant (ROA): An individual that has been certified by successful completion of an in house training program, to assist the licensed reactor operator during reactor operation.

1.19 Reactor Building: The Reactor Building includes the Reactor Bay, Control Room, and the Mechanical Equipment Room.

1.20 Reactor Safety System: Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

1.21 Reactor Secured: The reactor is secured when:

a.

There is insufficient fissile material or moderator present in the reactor core, adjacent experiments or control rods, to attain criticality under L

i optimum available conditions of moderation and reflection, or L

b.

The following conditions exist:

1.

All scrammable neutron absorbing control rods are fully inserted, and ii.

The Reactor Keyswitch is in the OFF position and the key is removed from the lock, and iii.

No work is in progress invoking core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and l-iv.

No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding one dollar (730 pcm).

1.22 Reactor Shutdown: That suberitical condition of the reactor where the absolute value of the negative reactivity of the core is equal to or greater than the shutdown margin.

1.23 Renortable Event: A Reportable Event is any of the following:

a.

Violation of a Safety Limit.

b.

Release of radioactivity from the site above allowed limits c.

Any of the following:

1.

Operation with actual Safety System Settings (SSS) for required systems less conservative than the Limiting Safety System Settings Amendment 11 3

December 15,1989

4 Anpendix A Technical Specincations (LSSS) specified in these specifications, ii.

. Operation in violation of Limiting Conditions for Operation (1f0) established in these Technical Specifications.

111.

A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown. (For components or systems other than those required by these Technical Specifications, the failure of the extra component or systems is not considered reportable provided that the minimum number of components or systems specified or -

- required perform their intended reactor safety function).

r iv.

An unanticipated or uncontrolled change in reactivity greater than one dollar (730 pcm). Reactor trips resulting from a known cause are excluded.

v.

Abnormal or significant degradation in reactor fuel, or cladding, or both; coolant boundary, or confinement boundary (excluding L

minor leaks), which could result in exceeding prescribed radiation j

exposure limits of personnel or environment, or both.

vi.

An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence of an unsafe condition with regard to reactor operations.

1.24 - Shim m4: A shim rod is a neutron absorbing rod having an in line drive which is mechanically, rather than magnetically, coupled and does not have a SCRAM capability.-

1.25 Shutdown Margin: Shutdown margin means the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by L

means of the control and safety systems starting from any permissible operating condition with the most reactive scrammable rod fully withdrawn, the non-scrammable rod (Shim rod) fully withdrawn, and with experiments considered at their most reactive condition, and finally, that the reactor will remain suberitical without further operator action.

1.26 True Value: The true value is the actual value of a parameter.

1.27 Unscheduled Shutdown : An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operations. This does not include shutdowns which occur during testing or check-out operations.

Amendment 11 4

December 15, 1989

i MPehdix A Technical Specifications 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS r

3 2.1 Safety Ilmits 2.1.1 Enfety thits (SL) for Forced Convection Flow.

Applicability This specification applies to the interrelated variables associated with core thermal and hydraube performance with forced convection flow. These interrelated variables are:

Reactor Thermal Power i

P

=

W Reactor Coolant Flow Rate

=

Reacto'r Coolant Inlet Temperature T.

=

Height of water above the top of the core H

=

Objective The objective is to assure that the integrity of the fuel clad is maintained.

Specification Under the condition of forced convection flow, 41 Safety Limit shall be as follows:

l a.

The combination of true values of reactor thermal power (P) and reactor coolant flow rate (W) shall not etceed the limits shown in Figure 2.11 i

under any operating conditions. The limits are considered exceeded if the point defined by the true values of P and W is at any time above the curve shown in Figure 2.11.

l b.

The true value of pool water level (H) shall not be less than 14 feet above the top of the core The true value of reactor coolant inlet temperature ( T

) shall not be c

greater than 120 F.

Balts Above 80 percent of the full core flow of 500 gpm in the region of full power operation, the criterion used to establish the Safety Limit was no bulk boiling at the outlet of any coolant channel. This was found to be far more limiting than the criterion of a minimum a!!owable burnout heat flux ratio of 2.0. The i

analysis is given in the SAR Appendix 3B.

Amendsment 11 5

December 15, 1989

Anpendis A Technical SpeclAcations In the region below 80 percent of full core flow, where, under a loss of flow transient at power the flow coasts down to zero, reverses, and then establishes natural convection, the criterion for selecting a Safety Umit is taken as a fuel cladding temperature. The analysis of a loss of flow transient is presented in Appenriix 3B of the SAR. For initial conditions of full flow and an operating power of 1.4 MWt, the maximum clad temperature reached under the conservative assumptions of the analysis was 273*F which is well below the temperature at which fuel clad damage could possibly occur. The Safety Limit i

shown in Figure 2.1-1 for flow less than 80 percent of full flow is the steady i

state power corresponding to the maximum fuel clad temperature of 273*F with natural convection flow, namely,1.4 MWt.

Annendment 11 6

December 18, 1989

ARDtadlLA 4

Technical Speelnentions l

2.1.2 Saferv unit (SIA for Natural Convection Flow.

Anplicability his specification applies to the interrelated variables associated with core thermal and hydraube performance with natural convection flow. The intenelated variables are:

P Reactor Thermal Power

=

T.

Reactor Coolant Inlet Temperature

=

H Height of water above the top of the core

=

Ohlective De objective is to assure that the integrity of the fuel clad is maintained.

Specification Under the condition of natural convection flow, the Safety Limit shall be as follows:

l a.

The true value of pool water level (H) shall not be less than 14 feet above the top of the core.

b.

The true value of reactor coolant inlet temperature ( T,.,,, ) sha!! not be greater than 120 F.

c.

The true value of reactor thenna! power (P) shall not exceed 1.4 MWt.

Balti The criterion for establishing a Safety Limit with natural convection flow is established as the fuel clad temperature. This is consistent with Figure 2.11 l

for forced convection flow during a transient. The analysis of natural convection flow given in Appendix 3B and 3C of the SAR shows that at 1.4 MWt the maximum fuel clad temperature is 273'F which is well below the temperature at which fuel clad damage could occur. The flow with natural convection at this power is 98 gpm. This flow is based on data from natural i

convection tests with fuel elements of the same design, performed in the I

prototype PULSTAR Reactor, as referenced in Section 3 of the SAR.

l Amendment 11 l

7 December 15,1989 1

I Aapendir A Techalcal Specincations 2.2 13mittne Safety System Settings l

2.2.1 11mitine Safety Svstem Settings (LSSS) for Forced Convection Flow Applicability

'Ihis specification applies to the setpoints for the safety channels monitoring reactor thermal power (P), coolant flow rate (W), and the height of water above the core (H).

Ohlective The objective is to assure that automatic protective action is initiated in order to prevent a Safety Umit from being exceeded.

i Specification Under the condition of forced convection flow, the Limiting Safety System Settings shall be as follows:

1.3 MWt (mat.)

P

=

W 450 gym (min.)

=

l H

14 feet, 2 inches (min.)

=

Bases The Umiting Safety System Settings that are given in the Specification 2.2.1 represent values of the interrelated variables which, if exceeded, shall result in automatic protective actions that will prevent Safety Umits from being exceeded during the most limiting anticipated transient (loss of flow). The safety margin that is provided between the Umiting Safety System Settings and the Safety Umits also allows for the most adverse combination of instrument uncertainties associated with measuring the observable parameters. These instrument uncertainties include a flow variation of ten percent, a pool level variation of two inches and a power level variation of seven percent.

'Ihe analysis presented in Section 3 of the SAR of a loss of flow transient indicates that if the interrelated variables were at their LSSS, as specified in 2.2.1 above, at the initiation of the transient, the Safety Umits specified in 2.1.1 would not be exceeded.

l Amendment 11 8

Dermber 15, 1989

Ag,tendix A Tichnical Specifkations 2.2.2 untilne Enfety System Settines (LSSS) for Natural convection Flow Applicability This specification applies to the setpoints for the safety channel monitoring reactor thermal power (P), and the height of water above the core (H).

Ohlective The objective is to assure that automatic protective action is initiated in order to prevent a Safety Umit from being exceeded.

Specifications Under the condition of natumi comection flow, the Limiting Safety System Settings shall be as follows:

P 250 kWt (max.)

=

H 14 feet, 2 inches (min.)

=

Bases The Umiting Safety System Settings that are given in Specification 2.2.2 represent values of the interrelated variables which, if exceeded, shall result in automatic protective actions that will prevent Safety Umits from being exceeded. The specifications given above assure that an adequate safety l

margin exists between the LSSS and the SL for natural convection. The safety i

margin on reactor thermal power was chosen with the additional consideration related to bulk boiling at the outlet of the hot channel. This criterion is not l

related to fuel clad damage (for these relatively low power levels) which was l

the criterion used in establishing the Safety Umits (see Specification 2.1.2). It is desirable to minimize to the greatest extent practical, 'N dose at the pool surface which might be aided by steam bubble rise during upflow in natural convection. Analysis of coolant bulk boiling given in SAR, Section 3, indicates that the large safety margin on reactor thermal power assumed in Specification 2.2.2 above will satisfy this additional criterion of no bulk boiling in any channel.

2 i

i Amendment 11 l

9 December 15,1989 I

1

1 Appendia A TechnicW Speelficatinas 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 M* actor r%ntant inlet Temnerature Anplicability his specification applies to the reactor coolant inlet temperature during operation with forced circulation and natural convection flow.

Objective De objective is to assure that the reactor will be operated within the bounds of established Safety Limits.

Specification The reactor coolant inlet temperature ( T

) as measured by either the Pool RTD or the Pool Thennal Switch shall not exceed 117'F.

Baats i

This specification requires that the coolant inlet temperature be mahtained below the maximum permissible measured value to assure that established Safety Limits and Umiting Safety System Settings are not affected. Coolant inlet temperature is a very slowly changing process variable that is monitored continuously. The marg, provided between the coolant inlet temperature m

alarm setting and the maximum permissible measured value of this process variable allows sufficient time for the operator to SCRAM the reactor without reaching the specified limit.

l Amendment 11 10 December 15,1989

Appendir A Technical Spaelncations 3.2 Reactivity Anplicability i

These specifications apply to the reacthity condition of the reactor and the 1

reactivity worths of control rods, shim rod and experiments.

Objective The objective is to assure that the reactor can be shut down at all times and that the Safety Limits will not be exceeded.

Speelfications The reactor shall not be operated unless the following conditions aist:

l a.

The shutdown margin, with the highest wonh scrammable control rod fully withdrawn, with the shim rod fully withdrawn, and with aperiments at their most reactive condition, relative to the Cold Critical condition, is greater than 400 pcm.

b.

The acess reactivity is not greater than 4070 pcm.

c.

The drop time of each control rod is not greater than 1.0 second.

d.

The rate of reactivity insenion of the control rods is not greater than 100 pcm per second (critical region only).

e.

The absolute reactivity wonh of aperiments or their rate of reactivity change shall not acced the values indicated in the following table:

Err >eriment Limit Movable 300 pcm or s 100 pcm/sec (whichever is more limiting)

Non secured 1000 pcm Secured 1700 pcm f.

The total reactivity wonh of all aperiments shall not be greater than 3000 pcm (sum of their absolute values).

Bases a.

The shutdown margin required by Specification 3.2a assures that the reactor can be shut down from any operating condition and will remain shutdown after cooldown and xenon decay, even if the highest worth scrammable rod should be in the fully withdrawn positionJ b.

The upper limit on excess reactivity ensures that an adequate shutdown margin is maintained.

Amendment 11 11 December 15, 1989

Anpendia A Technical EpmelAcations c.

The rod drop time required by Specification 3.2c assures that the Safety Umit will not be exceeded durm, g the flow reversal which occurs upon loss of forced convection coolant flow. The rise in fuel temperature due to heat storage is partially controlled by the reactivity insertion associated with the SCRAM. The analysis of this transient is based upon this SCRAM reactivity insertion taking the form of a ramp function of two second duration. This analysis is found in SAR Section 3.2.4 and Appendix 3B. The rod drop time is the time interval measured between the instant of a test signal in ut to the SCRAM Logic ' Unit and the instant of the rod seated si al.

d.

'Ihe maximum rate of reactivity insertion by the control rods which is I

allowed by Specification 3.24 assures that the Safety Umit will not be exceeded during a startup accident due to a continuous linear reactivity insertion. Refer to SAR Section 13.

Experiments affecting the reactivity condition of the reactor are commonly I

categorized by the sign of the reactivity effect produced by insertion of the experiment. An experiment having a large reactivity effect of either sign can also produce an undesirable flux distribution that could affect the peaking factor used in the Safety Umit calculations and the calibration of safety

channels, e.

The Specification 3.2e is intended to prevent inadvertent reactivity changes during reactor operation caused by the insertion or removal of an experiment. It further provides assurance that the failure of a single expenment will not result :n a reactivity insertion which could cause the Safety Umit to be exceeded. Analyses indicate that the inadvertent reactnity insertion of these magnitudes will not rcsult in consequences greater than those analyzed, SAR Sections 3.2.4,13.2.2.2 and 13.2.2.4.

f.

The total limit on reactivity associated with experiments ensures that an adequate shutdown margin is maintained.

Amendment 11 12 December 15, 1989

Appendia A Technical Specincations i

3.3 Reactor Safety System Applicability Dese specifications apply to the reactor safety system channels.

Objective De objective is to require the minimum number of reactor safety system i

channels which must be operable in order to assure that the Safety Umits are i

not exceeded.

Specifications The reactor shall not be operated unless the reactor safety.n' stem channels described in the following table' are operable:

Menwring Channel Function a.

Startup Power Levelm Inhibits Control Rod withdrawal when neutron count is s 2 cys.

b.

Safety Power Level SCRAhi at s 1.3 hiH' (LSSS), Enable for flow / flapper SCRAhiS at s 250 kH'(LSSS).

c.

Linear Power Level SCRAhi at s 1.3 AfH' (LSSS).

d.

Log N Power Level Enable for flow / flapper SCRAhiS at s 250 kU' (LSSS).

e.

Flow hionitoring*

SCRAAi when flapper not closed and flow / flapper SCRAhis are enabled, f

Primary Coolant Flow

• SCRAhi at n 450 gym (LSSS) when flow / flapper SCRAhis are enabled.

g.

Pool H'ater Temperature Alarm and hianual SCRAhi at s 117*F Monitoring Switch (LSSS).

h.

Pool H'ater Temperature Alann and Afanual SCRAhi at s 117*F Afeaturing Channel (LSSS).

L Pool H'ater Level SCRAhi at s 14 feet 2 inches.

J.

Afanual Button Afanual SCRAhi k

Reactor Keyswitch Afanual SCRAh!

Amendment 11 13 December 15, 1989 l

Appendix A Technical Speciflentions L.

Over the Poo!*

Alarm and Manual SCRAM Radiation Monitor

" Required only for reactor startup when power level is less than 4 watts.

• Either the Flapper SCRAM or the Flow SCRAM may be bypassed during maintenance testing and/or performance of a stanup checklist in order to

't verify each SCRAM is independently operable. The reactor must be shutdown in order to use these bypasses.

• Bypassed for less than two minutes during retum of a pneumatic rabbit capsule pom the core to the unloading station or five minutes during removal of experiments from the reactor pool.

U l

The Startup Channel inhibit function assures the required startup neutron source is sufficiem and in its proper location for the reactor startup, such that a minimum source multiplication count rate level is being detected to assure adequate information is available to the operator.

The reactor power level SCRAMS provide the redundant protection channels to assure that, if a condition should develop which would tend to cause the reactor to operate at an abnormally high power level, an immediate automatic protective action will occur to prevent exceeding the Safety Limit.

The alarms on the redundant pool water temperature channels are required to l

ensure that the reactor operator has adequate time to take corrective measures in order to prevent the pool temperature from reaching the Safety Limit.

l The primary coolant flow SCRAMS provide redundant protection 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 i

reactor will occur to prevent exceeding 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 recctor is above 250 kW (LSSS).

The pool water level channel together with the Over-the Pool (Bridge) radiation monitor, provides two diverse channels for shutdown of the reactor and prevents exceeding the Safety Limit due to insufficient pool height.

To prevent unnecessary initiation of the evacuation and confinement systems 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.

'Ihe manual SCRAM button and 'the Reactor Keyswitch provide two manual SCRAM methods to the reactor operator if unsafe or abnormal conditions should occur.

l.

I Amendment 11 1'

14 December 15,1989 l

l

Appendix A Techalcal Specincations 3.4 Reactor Instrumentation Applicability These specifications apply to the instrumentation that shall be available to the reactor operator to support the safe operation of the reactor, but are not considered reactor safety systems.

Objective The objective is to require that sufficient information be available to the operator to assure safe operation of the reactor.

Specifications Ihe reactor shall not be operated unless the following channels / systems / components listed are operable:

a.

"N Power Measuring Channel"'

b.

Control Rod Position Indications (for each Control Rod and the Shim rod) c.

Differential pressure gauge for " Bay with respect to Atmosphere"

'" Required when reactor power is greater than 500 kW.

Bases l

The "N Channel provides the necessary power. level information to allow i

adjustment of Safety and Linear Power Channels.

Control rod position indications give the operator information on rod height necessary to verify shutdown margin.

The differential pressure gauge provides the pressure difference between the Reactor Bay and the outside ambient and confirms air flow in the ventilation stream for both normal and confinement modes.

Amendment 11 15 December 15,1989

Anpendir A Technkal Spadftcations 3.5 Radiation Monitorine Eautoment Applicability

• nis specification applies to the ayallability of radiation monitoring equipment which must be operable during reactor operation.

Ohlective j

To assure that radiation monitoring equipment is available for evaluation 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.

a.

Threc pxed area monitors operating in the Reactor Building with their setpoints as follows:**

AIFRT SETPOINT AL4RM SETPOINT L

Control Room s 2.5 mR/hr s 25 mR/hr il Over the Pool s 10 mR/hr s 100 mR/hr lii.

West Wall s 10 mR/hr s 100 mR/hr b.

Paniculate and gas building ahaust monitors continuous {* sampling air in the facility ahaust stack with their setpoints as follows:m' AIFRT SETPOINTALARM SETPOINT L

Stack Gas s 200 Ar AfPC s 1000 *Ar hfPC il Stack Particulate s 200 '"Te AfPC z 2000 '"Te hfPC c.

The Radiation Rack Recorder.*

• For periods of time, not to aceed ninety days, for maintenance to the radiation monitoring channel, the intent of this specifcation will be satisfied if one of the installed channels is replaced with a gamma sensitive l

instrument which has its own alarm or is observable by the reactor operator or reactor opera!Or assistant.

• The Over the Pool bionitor may be bypassed for less than two minutes during retum of a pneumatic rabbit capsule from the core to the unloading 1

station or fee minutes during removal of aperiments from the reactor pool.

• hiay be bypassed for less than one minute immediately after staning the Amendment 11 16 December 15, 1989

i n

t Anpandix A Technical Specifications pneumatic bloner system.

"7he setpoints for the Stack Gas and Paniculate are based on the MPC quantities present in the ventilation flow stream as it crits the stack.

l

• During periods of repair and/or maintenance of the recorder, the specified area and effluent monitor's readings shall be recorded manually at a nomina! interval of 30 minutes when the reactor is not shutdown.

Bassa 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 SAR.

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, or stack exhaust monitor, the building will be automatically placed in confinement as described in Section 5 of the SAR.

To prevent unnecessary initiation of the evacuation confm* ement 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 during the specified time interval, 1

I Amendment 11 17 Dmmber ts,1989

1 Appendin A i

Tecknami Speelfications 3.6 Connnement and Main HVAC Systems Annlicability his specification applies to the operation of the Reactor Building confinement and main HVAC systems.

Ohlective De objective is to as:;ure that the confinement system is in operation to mitigate the consequences of possible release of radioactive materials resulting from reactor operation.

Speelnention 7he reactor shall not be operated, nor shall irradiated fuel be moved within the pool area, unless the folloning equipment is operable, and conditions met:

EatapmentNondition Function a.

All doors, except the Control To maintain reactor building Room and basement corridor entrance negative differential pressure selflatching, self closing, closed (dy).*

and locked.

i b.

Control room and basement corridor To maintain reactor building entrance door; selflatching, negative differential pressure.*

self closirg and closed.

c Reactor Building under a negative To maintain reactor building differential pressure of not less negative differential pressure uith than 0.2" H,0 n'ith the normal reference to outside ambient.*

ventilation system or 0.1" H,0 uith l

one confnement fan operating.

d.

Confinement system Operable""*

l e.

Evacuation system Operable *'

f R 3 Ventilation Fans Operable l

• Doors may be opened by authorized personnel for less than Svc minutes for personnel and equipment transport provided audible and visual indication is available for the reactor operator to verify door status.

• Doors may be opened for periods of less than pre minutes for personnel and equipment transport between corridor crea and Reactor Building.

• During an interval not to etceed 30 minutes after a loss of dp (nith Main l

Amendment 11 18 December 15,1989 m

)

Annandix A Technical SpadReatinas HVAC operating) is identiped, reactor operation may continue while the

+

loss of dp is inn:stigated and corrected.

" Operability also demonstrated uith an auxiliary poner source.

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• One fiter train,nay be out of sen' ice for the purpose of maintenance, sepair, and/or surveillance Durirng the period of time mlor a period of time not to exceed 45 day standby fiter train shall be operating with the Reactor Building in the i

confnement mode.

"The public address system can serve temporarily for the Reactor Building evacuation system during short periods of maintenance.

Bases In the event of a fission product release, the confinement initiation system will secure the normal ventilation fans and close the normal inlet and exhaust dampers. In confinement, a confinement fan system will: maintain a negative pressure in the Reactor Building and insure in leakage only; purge the air from the building at a greatly reduced and controlled flow through charcoal and absolute filters; and control the discharge of all air through a 100 foot stack on site. Section 5 of the SAR describes the confinement system's sequence of operation.

'Ihe allowance for operation under a temporary loss of dp when in normal ventilation is based on the requirement of having the confinement system operable and therefore ready to respond in the unlikely event of an airborne release.

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Amendment 11 19 December 15,1989 1

i Appendix A j

Tarhalcal Specificatlans 3.7 umitatlans or Exneriments i

Applicability i

his specification applies to experiments installed in the reactor and its 1

experimental facilities. Fueled experiments must also meet the requitements of Specification 3.8.

Objective De objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications

'Ihe reactor shall not be operated unless the following conditions goveming aperiments exist:

\\

a.

All materials to be irradiated shall be either corrosion resistant or encapsulated within a corrosion resistant container to prevent interaction with reactor components or pool water. Corrosive materials shall be doubly encapsulated.

i b.

Irradiation containers to be used in the reactor, in which a static pressure will aist or in which a pressure buildup is predicted, shall be designed and tested for a pressure aceeding the marimum expected by a factor of 2.

Pressure buildup inside any container shall be limited to 200 psi.

c.

Cooling shall be provided to prevent the surface temperature of an experiment to be irradiated from exceeding the saturation temperature of the reactor pool water.

l d.

Experimental apparatus, material or equipment to be inserted in the reactor l

shall be positioned so as to not cause shadowing of the nuclear instrumentation, interference with control rods, or other perturbations which may interfere with safe operation of the reactor.

i e.

Conceming the material content of experiments, the following will apply:

i.

No aperiment will be performed unless the major constituent content of the material to be irradiated is known and a reasonable effort has been made to identify trace elements and impurities whose activation i

may pose the dominant radiological ha:ard. H' hen a reasonable effort does not give conclusive information, one or more short irradiations of small quantitics of material may be pe.rformed in order to identify the activated products.

ii.

Attempts will be made to identify and limit the quantities of elements having ver,r Lg:. wmal neutron absorption cross sections, 1

Amendment 11 20 Deember 1$,1989 1.

4pendlz A Technical Specifications i

in order to quantify reacthity effects.

lii.

Explosive material"', shall not be allowed in the reactor.

Experiments reviewed by the Radiation Protection Council in which the material h considered to be potentially explosive, either while contained, or if it leaks )>om the container, shall be designed to maintain seal integrity even if detonated, to prevent damtge to the reactor core or to the control rods or instrumentation and to prevent any change in reacthity, iv.

Each experiment will be evaluated with respect to radiation induced physical and/or chemical changes in the irradiated material, such as decomposition effects in polymers.

v.

Erperiments im>oMrng pammable"' or highly toxic materials"' require specipc procedures for handling and shall be limited in quantity as approved by the Radiation Protection Council. No cryogenic liquids"' will be allowed uithin Ihe biological shield of Ihe PULSTAR Reactor.

g.

Credible failure of any erpetiment shall not result in releases or erposures in excess of the annuallimits established in 10 CFR 20.

"'Depned as follows (reference National Fire Code omless othernise noted):

Toxic:

A substance that has the ability to cause damage to living tissue when inhaled, ingested or absorbed via the skin.

Flammable: Having a flash point below 100*F and vapor pressurc l

not exceeding 40 psia. The flash point is depned as the minimum temperature of a liquid at which suffcient vapor h given off to form an ignitable mixture with the air near the surface of the liquid or within the vessel used as determmed by appropriate test procedures and apparatus as speciped.

Explosive:

Any chemical compound, mixture, or device, the primary or common purpose of which is to function by explosion. The term includes, but is not limited to, dynamite, black ponder, pellet ponder, initiating explosives, detonators, safety fuses, squibs, detonating cord, igniter cord, and igniters.

{

Cryogenic:

A cryogenic liquid is considered to be a liquid with a normal boiling point below 238cF (reference -

National Bureau of Standards Handbook 44).

Amendment 11 21 December 15,1989 i

a Appendix A Technical Specifications Ranta Specifications 3.7a, 3.7b, 3.7c, and 3.74 are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from expermment failure; and, serve as a guide for the review and approval of new and untried experiments by the facility personnel, as well as the Radiation Protection Council.

Specification 3.7e insures that no physical or nuclear interferences compromise the safe operation of the reactor, specifier,lly, an er>eriment having a large reactMty effect of either sign could produce an uncestrable flux distribution that could affect the peaking factor used in the Safety Umit calculation and/or safety channels calibrations. Review of the experiments using these ILOs and the Administrative Controls of Section 6 will insure the insertion of experiments will not negate the considerations implicit in the Safety Umits and thereby become an Unreviewed Safety Question.

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l Amendment 11 22 Domber 15,1989

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i Appendix A Technical Epacificatlans 3J pneration with Fueled Erneriments l

f Applicability This specification applies to the operation of the reactor with any fueled experiment.

Ohlective To assure that the confinement leak rate and fission product inventory are 4

within the limits used in the PULSTAR Safety Analysis and are consistent with i

i present U. S. Nuclear Regulatory Commission guides and the Code of Federal Regulations.

Specifications The following limitations apply to fueled aperiments at the PULSTAR Reactor:

a.

For fueled experiments operated in the reactor pool:

i.

Dse aperiment is under at least 18 feet of water.

li.

7he thermal power (or pssion rate) generated in the aperiment is not greater than 50 watts (1.6 x 10' pssions/sec).

1 iii.

The total aposure of fueled aperiment material in the reactor is not greater than 2.3 x 10,* pssions (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 50 watts).

iv.

No more than 110 milligrams of *"U will be used in the aperiment.

b.

For fueled aperiments in reactor aperimental facilities other than within the reactor pool:

1.

The thermal power (or pssion rate) generated in the aperiment is not greater than 1 watt (3.2 x 10* pssions per second).

il The total aposure of the material is not greater than 2.8 x 10' fssions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at I watt).

l c.

The reactor shall not be operated with a fueled aperiment unless the ventilation is operated in the confnement mode.

d.

The specifcations pertaining to reactor aperiments, detailed in Section 3.7 Limitations of Experiments, apply.

Baats in the event of the failure of a fueled experiment with the subsequent release of fission products (25% of iodine isotopes and 100% of noble gas isotopes)

Amendment 11 23 pensber 15, 1989

1

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Anmadh &

i; f ach alant s pa d a m u m the 30 day inhalation exposure to these isotopes at any location is less than the l

limits set by 10 CFR 20 using the averaging period of one year, In making the safety analysis, the assumptions used for evaluating the potential radiological

! =.

consequences of a failed fueled experiment conformed to those of Nuclear Regulatory Commission Regulatory Guide 1.3.

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Amendment 11 l-24 December 15,1989

i Ampendia A Tackalcal EymelReatlans 3.9 Primary Coolant Applicability These specifications apply to the water quality and flow path of the pnmary coolant.

l objective The objective is to require that primary coolant cuality be maintained to acceatable values in order to reduce the potentia'. for corrosion and limit the i

built up of activated contaminants in the primary piping and pool.

Specifications The reactor shall not be opera ed $nless the pool water meets the following limits:

a.

The resistivity shall be a 500 kn.cm.

l b.

The pH shall be within the range of S.S to 7.5.

l c.

Core grid plate holes must be plugged whenever the reactor is not shutdown and the primary coolant is in forced convection.

Balts i

The limits on resistivity are based on reducing the potential for corrosion in the primary piping or pool liner and to reduce the potential for activated contaminants in these systems.

Specification 3.9c ensures that flow safety limits and limiting safety system settings are applicable to only flow through the fuel assemblies.

Amendment 11 25 December 15, 1989

Annendla A Technleni Sngincations 4.0 SURVEILLANCE REOUIREMENTS The intent of the sun'eillance interval (e.g., annually, but not to exceed 15 months) is to maintain an average cycle, with occasional extensions as allowed by the interval tolerance.

If it is desired to permanently change the scheduled date of a sun'eillance (e.g.,

i move an August scheduled date to an April scheduled date because of lower seasonal temperatures), the particular surveillance item will be performed at the earlier date and the associated intenal normalized to this revised earlier date. In no cases will permanent scheduling changes (that yield slippage of the surveillance interval's routine scheduled date) be made by using the allowed intenal tolerance.

t

' Surveillance tests, (except those specifically required for safety when the reactor is shutdown) may be deferred during reactor shutdown; however, they must be completed prior to reactor startup (or immediately after the startup, if and only if reactor operations are required to perform the surveillance item).

Suncillance requirements scheduled to occur during operation which cannot be performed with the reactor operating may be deferred until planned reactor shutdown.

l.1 Eu.cl Applicability This specification applies to the surveillance requirement for the reactor fuel.

Oblective The objective is to monitor the physical condition of the PULSTAR fuel.

Specincations All fuel assemblics shall be visually inspected biennially but at intervals not to exceed thiny (30) months.

Bases 1

The assemblies are inspected for physical damage including corrosion of endplates, nosepiece and zircaloy box; missing assembly screws; dented and scratched surfaces; and blockage of coolant channels.

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The biennial inspection of PUIJSTAR fuel assemblies in conjunction with the monthly primary coolant analysis has been shown to be adequate for Zr 2 clad assemblies to insure fuel assembly integrity based on a long history of the prototype PULSTAR steady state and pulse operation.

Amendment 11 26 December 15,1989 i,-

~.

Appendia A Tachaical SpaclRentions 4.2 Contml Rods Applicability This specification applies to the surveillance requirements for the control rods, shim rod, and control rod drive mechanisms (CRDM).

Ohlective The objective is to assure the operability of the control rods and shim rod, and to provide current reactivity data for use in verifying adequate shutdown margm.

Specifications The reactivity worth of the s'him rod and each control rod shall be a.

detennined annually but at inten'als not to aceed ffteen months for the i

steady state core in current use. The reactivity worth of all rods shall be determined for any new core or rod confguration, prior to routine operation.

b.

Control rod drop times"' and control rod drive times shall be determined:

(1) annually but at intenals not to aceed ffteen months, and (2) after a control assembly is moved to a new position in the core or after maintenance or modifcation is perfonned on the control rod drive mechanism.

c.

The control rods shall be visually inspected biennially but at intervals not to aceed thirty months.

d.

The values of acess reactivity and shutdown margin shall be determined monthly, but at intervals not to aceed six weeks.

• Applies only to magnetically coupled rods.

Balti The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide a means for determining the reactivity worths of experiments inserted in the core. 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.

Control rod drive and drop time measurements are made to determine whether the rods are functionally operable. These time measurements may also be utilized in reactor transient analysis.

Amendment 11 27 rmember ts,1989

Appendix A Tachalmal SandAmtlann Visual inspections include: detection of wear or corrosion in the rod drive mechanism; identification of deterioration, corrosion, flaking or bowing of the -

neutron absorber material; and verification of rod travel setpoints.

Control rod surveillance procedures will document proper control rod system i

reassembly after maintenance and recorded post maintenance data will identify signincant trends in rod performance.

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L Amendament 11 28 December 15,1989

l Appendix A Technical Spadficatleta 4,3 Itemeter Instrumentation and Safety Systems Applicability his specification applies to the surveillance requirements for the Reactor Safety System and other required reactor instrumentation,.

Objective De objective is to assure that the required instrumentation and Safety Systems will remain operable and will prevent the Safety Limits from being exceeded, t

Specifications n.

A channel check of 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 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 "N Channel shall be made semi annually, but at intervals not to exceed seven and one half months. A calorimetric measurement shall be performed to determine the "N detector current associated with full power operation.

d.

A channel calibration of the following channels shall be made semi-annually but at intervals not to exceed seven and one half months. "'

1.

Pool Water Temperature (T) 2.

Primary Cooling and Flow Monitoring (Flapper) 3.

Pool Water Level 4.

Primary Heat Exchanger Inlet (TJ and Outlet Temperature (TJ 5.

Safety and Linear Power Channels

"'A channel calibration shall also be required after repair of a channel

[

component that has the potential of affecting the calibratiort of the channel.

l Baats ne daily channel tests and checks will assure that the Reactor Safety Systems are operable and will assure operations within the limits of the operating license. The semi annual calibrations will assure that long term drift of the channels is corrected. The calorimetric calibration of the reactor power level, Amendment 11 29 December 15, 1989

g_

Ammandis A Tm EN L

in cordunction with the Nitrogen 16 Channel, provides a continual reference for y

adjustment of the Linear, Log N and Safety Channel detector positions.

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^"'*d*'"' 11 30 December 15, 1989

f.'

ARMadlLA Techalcal Epecincations 4.4 Radiation Monitorine Eaulnment Applicability This specification applies to the surveillance requirements for the area and stack effluent radiation monitoring equipment.

Ohlective The objective is to assure that the radiation monitoring equipment is operable.

Specification (a)

The area and stack monitoring systems shall be calibrated annually but at intervals not to exceed pfteen months.

(b)

The setpoints shall be veriped weekly, but at intervals not to exceed 10 days.

Baits 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 Amendment 11 l

31 December 15,1989 1

7 I

?

Appendia A Technical Epmeincations J

4.5 Connnement and Main HVAC System Anplicability

(

This specification applies to the surveillance requirements for the confinement and main HVAC systems.

t Objective The objective is to assure that the confinement system is operable.

?

Specifications a.

The confnement and evacuation system shall be veriped to be operable within seven days prior to reactor operation, b.

Operability of the confnement system on auxiliary power will be checked monthly but at inten'als not to exceed six neeks.*

c.

A visual inspection of the door seals and closures, dampers and gaskets of the confnement and ventilation systems shall be performed semi annually but at inten>als not to exceed seven and one half montlu to verify they are operable.

d.

The control room differential pressure (dp) gauges shall be calibrated annually but at intervals not to exceed pfteen months.

l e.

The conpnement pher train shall be tested biennially but at intervals not to exceed thirty months and prior to reactor operation following confnement HEPA or carbon adsorber replacement. This testing shall include iodine adsorption, particulate removal effciency and leak testing of the fiter housing.*

f.

The air flow rate in the confnement stack exhaust duct shall be detennined annually but at inten'als not to exceed pfteen months.

g.

The air fow rate of the Burl lngton South Wing (R 3) ventilation fans shall be determined annually, but at intervals not to exceed ffteen months.

• Operation must be verified following modifcations or repairs invoh'ing load changes to the auxiliary power source.

• Testing shall also be required following major maintenance of the fiters or housing.

Bases Surveillance of this equipment will verify that the confinement of the Reactor Building is maintained as described in Section 5 of the SAR.

Amendment 11 32 December 15, 1989

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P Apnendix A Technical EneclAct.tlons 4.6.

Primarv and Secondary Coolant Appilcability This specification applies to the surveillance requirement for monitoring the radioactivity in the primary and secondary coolant.

Ohlective

- The objective is to monitor the radioactivity in the pool water to verify the integrity of the fuel cladding and other reactor structural components. The secondary water analysis is used to confirm the boundary integrity of the primary beat exchanger.

Specifications The primary coolant shall be analyzed monthly, but at intervals not to a.

exceed sh weeks. The analysis shall include gross beta / gamma counting, gamma spectroscopy, Neutron Activation Analysis (NAA), pH and resistivity measurements.

b.

The secondary coolant shall be analyzed monthly, but at intervals not to exceed sh weeks. This analysis shall include, as a minimum, gamma spectroscopy.

Baits Radionuclide analysis of the pool water samples will allow detection of fuel clad failure, while neutron activation analysis will give corrosion data associated with primary system components in contact with the coolant.

The detection of radioactivity (above backgrcund) in the secondco coolant provides evidence of a primary heat exchanger leak.

Amendment 11 33 December ts,1989

Appendir A i

Technical Speelneations 5.0

])ESIGN FEATURES 5.1-E p.asto t E ucl a.-

The reactor fuel shall be UO, with a nominal enrichment of 4% in *"U,.

zircaloy clad, with fabrication details as described in Section 3 of the

. Safety Analysis Report.

b.

Total burnup on the reactor fuel is limited to 20,000 MWD /MTU.

5.2 Reactor Building

a. -

The reactor shall be housed in the Reactor Building, designed for confinement. The minimum free volume in the Reactor Building shall be 2.25 x 10P cm' (refer to SAR Chapter 13 analysis).

o

b. -

The Reactor Building ventilation and confinement systems shall be separate from the Burlington Engineering Laboratories building systems and shall be designed to exhaust air or other gases from the building E

through a stack with discharge at a minimum of 100 feet above ground L

level, o

l-c.

The openings into the Reactor Building are the truck entrance door, personnel entrance doors, and air supply and exhaust ducts, d.

The Reactor Building is located within the Burlington Engineering Laboratory complex on the north campus of North Carolina State 1.

University at Raleigh, North Carolina. The restricted access areas of the facility include the PULSTAR Control Room, Reactor Bay, Mechanical 1

Equipment Room, nnd Waste Tank vault.

5.3 Fuel Storage Fuel, including fueled experiments and fuel devices not in the reactor, shall be stored in a geometrical configuration where k,, is no greater than 0.9 for all conditions of moderation and reflection using light water except in cases where a fuel shipping container is used, then the k,e for the container shall apply, 5.4 Remetivity Control Reactivity control is provided by four neutron absorbing blades. Each control

~ blade is nominally comprised of 80% silver,15% indium, and 5% cadmium with nickel cladding. Three of these neutron absorbing blades are magnetically coupled and have scramming capability. The remaining neutron. absorbing blade is non-scrammable. One of the scrammable rods may be used for automatic servo control of reactor power. When in use, the servo-control maintains a constant power level as indicated by the Linear Power Channel.

Amendment 11 34 December 15, 1989

p-i; Apasadin A i

Technical Specifications 5.5 Prl=m Coolant System De Primary Coolant System consists of the aluminum lined reactor tank, a "N

' decay tank, a pump, and heat exchanger, and associated stainless steel piping.

De nominal capacity of the primary system is 15,600 gallons. Valves are located adjacent to the biological shield to allow isolation of the pool, and at myor components in the primary system to permit isolation.

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Amendment 11 35 December 15,1989

' Anatadits Irrhnical Snecifications (i.0

ADMINISTRATIVE CONTROLS 6.1 Oreanization The reactor facility shall be an integral part of the Department of Nuclear Engineering of the College of Engineering of North Carolina State University.

The reactor shall be related to the University structure as shown in Figure 6.11. Responsibility for the safe operation of the PULSTAR Reactor shall be with'the chain of command established in Figure 6.1-1. Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be responsible for safeguarding the

_ public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license and these technical specifications.

6.1.1 Ornanizational Structure The following specific organizational levels (as defined by ANSI 15.11982) and positions shall exist at the PULSTAR Facility:

LEVEL 1: This level shall include the Chancellor, the College of Engineering Dean, the Nuclear Engineering Department Head, and the Director of the Nuclear Reactor Program.

The Director of the Nuclear Reactor Program is responsible for the long range development of the Nuclear Reactor Program and for the general conduct.of Program operations. He evaluates new service and research applications for the PULSTAR Reactor and Scaled PWR Facility, recruits new users of the facilities, supervises the development and expenditure of Program budgets, develops department and university support for needed capital investment, and works through the Associate Director to monitor daily operations. While the Associate Director supervises daily operations and makes routine decisions related to safety and operation, the Director is responsible for reactor performance and personnel matters.

LEVEL 2 - Associate Director of the Nuclear Reactor Procram: The Associate Director is responsible for the safe and efficient operation of the PULSTAR Reactor Facility. In matters pertaining to the operation of the facility and these Specifications, the Associate Director reports to the Director of the Nuclear Reactor Program. The Associate Director and Director of the Nuclear Reactor Program consult with the Head, Department of Nuclear Engineering on PULSTAR operations matters as required. The minimum qualifications for the Associate Director are at least five years of reactor operating experience including at least two years of supervisory reactor experience. Baccalaureate or graduate study may be substituted for a maximum of one year reactor operating experience.

LEVEL 3 - Reactor Onerations Mananer: The Reactor Operations Manager, who shall be qualified as a Class A Operator, shall be responsible for assuring that operations are conducted in a safe manner and within the limits prescribed Amendment 11 36 Decemher is,1989

D 1

Appendis A I

Technical Specifications

,by the facility license, all applicable Nuclear Regulatory Commission j

. regulations, and the provisions of the Radiation Protection Council. The-o Reactor Operations Manager reports directly to the Associate Director of the Nuclear Reactor Program.

IFVEL 4. Onerating Staff: This level includes the positions of Chief Reactor b

Operator, Chief of Reactor Maintenance, and the remaining Class A and B operators. Personnel at this level report to the Reactor Operations Manager s

(for PULSTAR Reactor related matters).

i Reactor Health Physicist: The Reactor Health Physicist is responsible for assuring the safety of reactor operations from the standpoint of radiation protection. The Reactor Health Physicist reports directly to the Nuclear Engineering Department Head and shall function independent of the campus Radiation Protection Office as shown in Figure 6.11. He shall possess relevant practical experience in the application of health physics principles.

In a!! instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon the appropriate qualipcations.

6.1.2 Minimum Stamng The minimum staffing when the reactor is not secured shall be:

a.

A certified reactor operator (either Class A or B) in the Control Room.

b.

' Reactor Operator Assistant (ROA), capable of being at the reactor facility within five minutes upon request of the reactor operator on duty.

When the ROA leaves the Burlington Engineering Laboratories complex, he shall carry a paging device such that he can be summoned back within the specified time limit. The ROA staff function may be filled by the Class A Reactor Operator (required by item c.). The unexpected absence of the ROA for up to two hours to accommodate a personal emergency is allowed, provided immediate action is taken to obtain a replacement.

- c.

A Class A Reactor Operator. This individual may be referred to as the

" Designated Senior Reactor Operator (DSRO)" and shall be readily on call, meaning:

i.

Has been specifically designated and the designation known to the reactor operator on duty.

ii.

Keeps the reactor operator on duty informed of where he may be rapidly contacted and the phone number.

iii.

Is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g.,30 minutes or within a 15 mile radius).

Amendment 11 37 December 15, 1989

Appendix A Technical Speelnentions d.

A Reactor Health Physicist, or his designated alternate. This indMdual shall also be on call, under the same limitations as prescribed for the Class A Reactor Operator under specification (c).

+

6.1.3 Class A Reactor Operator Duties' The following events shall require the presence of a Class-A Reactor Operator at the Burlington Engineering Laboratory complex:

a.

Initial startup and approach to power (i.e., required after each time interval that the reactor is secured),

b.

All fuel or control rod relocations within the reactor core or pool, c.

Relocation of any in core experiment with a reactivity worth greater than one dollar, d.

Recovery from unplanned or unscheduled shutdown or significant power reduction (documented verbal concurrence from a Class A Reactor Operator is acceptable).

Amendment 11 38 December 15,1989

m

~ Anw ndix A Technical Specifications 6.2 Review and Audit 6.2.1 - Radiation Protection Council There shall be a Radiation Protection Council (RPC) whose duties shall 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 RPC Comnosition and Oualifications a.

The Radiation Protection Council shall have a minimum of five members, of whom less than a majority shall be from the line organization shown in Figure 6.11. In addition to the Radiation Protection Council, a minimum of three persons are appointed by the Chancellor, upon the recommendation of the RPC, to form a Reactor Safeguards Advisory Group (RSAG). This group serves as a permanent committee to the RPC and will be solely responsible for an independent audit of reactor operations. The RPC and RSAG shall be made up of faculty and staff who shall collectively provide experience in reactor engineering, reactor operations, chemistry and radiochemistry, instrumentation and control systems, radiological safety, materials and mechanical and electrical systems. The campus Radiation Protection Officer (RPO) shall be a permanent member of the RPC. The RPC shall prescribe which review items (detailed under 6.23) that are to be delegated to the RSAG.

b.

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.

c.

The RPC shall meet annually, but at intervals not to exceed fifteen months and upon call of the Chairman; while the RSAG shall meet as needed or as specifically required by the audit function detailed under Technical Specification 6.2.4 or upon call of the Chairman.

6.23 RPC/RSAG Review Function The following items shall be reviewed by the RPC (and as needed by referral to RSAG):

a.

Determinations that proposed changes in equipment, systems, test, experiments, or procedures do not involve an unreviewed safety question.

b.

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

c.

All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity.

Amendment 11 39 December 15, 1989

L.

(

Apnendix A Technical Specifications I

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Proposed changes to the Technical Specifications or facility license.

c.

Violations of technical specifications or license. Violations of internal procedures or instructions having safety significance.

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f.

Operating abnormalities having safety significance.

g.

Reportable Events (as per technical specification definition 1.24).

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h.

Audit reports.

l A written report of meeting minutes and the findings or recommendations of j

the review group shall be submitted to Nuclear Engineering Department Head, i

Director of the Nuclear Reactor Program, Associate Director of the Nuclear

- Reactor Program, the Radiation Protection Council members and the Reactor i

Safeguards Advisory Group, in a timely manner after the review has been completed.

l 6.2.4 RSAG Audit Function The audit function shall consist of selective (but comprehensive) examination of i

operating records, logs, and other documents. Discussions with cognizant

-i personnel and observation of operations shall also be used as appropriate. The Reactor Safeguards Advisory Group (RSAG) under the authority of the i

l Radiation Protection Council (RPC), shall be responsible for this audit function. The audit shall include:

a.

Facility operations for conformance to the technical specifications and license, annually, but at intervals not to exceed fifteen months.

L b.

The retraining and requalification program for the operating staff, biennially, but at intervals not to exceed thirty months.

c.

The results of action taken to correct those deficiencies that may occur l

in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety, annually, but at intervals not to exceed fifteen months.

d.

The Emergency Plan and Emergency Procedures, biennially, but at intervals not to exceed thirty months.

Depciencies uncovered that affect reactor safety shall be immediately reported to the Nuclear Engineering Department Head, Director of the Nuclear Reactor Program and the Associate Director of the Nuclear Reactor Program. A written report of the findings of the audit shall be submitted to the Nuclear Engineering Department Head, the Director of the Nuclear Reactor Program, the Associate Director of the Nuclear Reactor Program, The Radiation Amendment 11 40 December 15, 1989

, o;;.

> :f Annendix A e

Technical Speelficatinas

= Protection Council and the Reactor Safeguards' Advisory Group members, within three months after the ' audit has been completed, s

t i.

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i Amendment 11 41 December 15,1989

~ _

-r Anpendix A

{

Techn Enl SpeelReations 6.3 Oneratino Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the following:

j a.

Startup, operation and shutdown of the reactor.

b.

Fuel loading, unloading, and movement within the reactor.

c.

. Maintenance of major components of systems that could have an affect on reactor safety.

d.

Surveillance checks, calibrations and inspections required by the technical specifications or those that may.have an affect on reactor safety.

Personnel radiation pro'tection, consistent with applicable regulations and c.

y that include commitment and/or programs to maintain exposures and releases as low as reasonably achievable (ALARA).

f.

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

g.

Implementation of the emergency plan and security plan.

Substantive changes to the above' procedures shall be made effective only after documented review by the RPC (or RSAG as applicable) and approval by the Associate Director of the Nuclear Reactor Program, or his designated alternate.

Minor modifications to the original procedures which do not change their original intent may be made by the Reactor Operations Manager, but the modifications shall be approved by the Associate Director of the Nuclear Reactor Program within 14 days.

Temporary deviations from procedures may be made by the Class A Reactor Operator (on duty as required by specification 6.1.2 c.) or Reactor Operations Manager, in order to deal with special or unusual circumstances or conditions.

Such deviations shall be documented and reported to the Associate Director of the Nuclear Reactor Program, or his designated alternate.

Amendment 11 42 December 15, 1989

~

Anaendix A l

Technical Speelfications 6A Review of Ernerimentti 6A.1 New (untried) Erneriments All new experiments or class of experiments, referred to as " untried" experiments, shall be reviewed and approved by the Associate Director of the Nuclear Reactor Program, Reactor Health Physicist, and the Radiation Protection Council (or RSAG as applicable), prior to initiation of the experiment.

The review of new experiments shall be based on the limitations prescribed by Technical Specifications 3.7 and 3.8 and other Nuclear Regulatory Commission regulations, as applicable, if the Radiation Protection Council, the Associate Director of the Nuclear Reactor Program, and the. Reactor Health Physicist jointly agree that the experiment can be safely performed within the limitations of the technical specificationsLand other applicable Nuclear Regulatory Commission regulations, then an approved PULSTAR Project Number can be issued by the RPC for the experiment.

6A.2 Tried Exneriments All proposed experiments are reviewed by the Reactor Operations Manager and the Reactor Health Physicist (or their designated alternates). Either of these individuals may deem that the proposed experiment is not adequately covered by the documentation / analysis associated with an existing approved PULSTAR Project and therefore constitutes an untried experiment that will require the approval process detailed under Technical Specification 6.4.1. If the Reactor Operations Manager and the Reactor Health Physicist concur that the experiment is a tried experiment, then the request is approved and the experiment can be scheduled within the limitations of the reactor operating schedule.

Substantive changes to previously approved experiments shall be made only after review and approval by the Associate Director of the Nuclear Reactor Program, Reactor Health Physicist, and the Radiation Protection Council (or RSAG as applicable).

Amendment 11 43 December 15,1989

p --

Appendix A 4

Technical Speelnentions

]

6.5 Action to be Taken in Case of Safety Limit Violation In the event a Safety Umit is violated.

a.

The reactor shall be shut down and reactor operations shall not be resumed until authorized by the Nuclear Regulatory Commission.

b.

The Safety Umit violation shall be promptly reported to the Associate Director of the Nuclear Reactor Program, or h.s designated alternate.

I c.-

The Safety.Umit violation shall be reported to the Nuclear Regulatory Commission in accordance with specification 6.7.1.

i d.

A Safety Umit violation report shall be prepared that describes the following: -

l.

i.

Circumstances leading to the violation including, when known, the l-cause and contributing factors.

11.

Effect of violation upon reactor facility components, systems, or structures and on the health and safety of facility personnel and l-l the public.

l l

iii.

Corrective action to be taken to prevent recurrence.

The report shall be reviewed by the Radiation Protection Council and any l

follow up report shall be submitted to the Nuclear Regulatory Commission l

when authorization is sought to resume operation.

l l

l l

l l

l l

I Amendment 11 l

44 December 15, 1989

y F

Appendix A Technical Epecincations

+

6.6 -

Action to be Taken for Renortable Events (other than SL Violation)

)

. In case of a Reportable Event (other than violation of a Safety Limit), as defined by section 1.23 of these specifications, the following action shall be taken:.

a.

Reactor conditions shall be returned to normal or the reactor shall be 1

shutdown. -If it is necessary to shutdown the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Associate Director of the Nuclear Reactor Program, or his designated alternate.

-l b.

The occurrence shall be reported to the Associate Director of the Nuclear Reactor Program, and to the Nuclear Regulatory Commission in l

L accordance with specification 6.7.1 r

c.

The occarrence shall be reviewed by the Radiation Protection Council at their next scheduled meeting.

c l'

i l

h L

Amendment 11 45 December 15,1989

P Appendir A Technical Specifications 6.7

' Renonine Reautrements 6.7.1 Reportable Event For Reportable Events as defined by section 1.24 of these specifications, there shall be a report not later than the following work day by telephone to the Nuclear Regulatory Commission Operations Center and the Nuclear Regulatory Commission Region II Regional Administrator, followed by a written report within 14. days that describes the circumstances of the event l

6.7.2 Permanent Changes in Facility Organization l

Permanent changes in the facility organization involving either 1.evel 1 or 2 personnel (refer to specification 6.1) shall require a written report within 30 days to the Nuclear Regulatory Commission Operations Center and the Nuclear Regulatory Commission Region II Regional Administrator.

i 6.7.3 Channes Associated with the Safety Analysis Renort Significant changes in the transient or accident analysis as described in the Safety Analysis Report shall require a written report within 30 days to the L

Nuclear Regulatory Commission Operations Center and the Nuclear Regulatory Commission Region II Regional Administrator.

L 6.7.4 Annual Oneratine Renort An annual operating report is required to be submitted no later than August

'i 31st of each year and will cover the period of July 1st through June 30th. The original is transmitted to the Document Control Desk, I4uclear Regulatory Commission, Washington, with a copy transmitted to the Nuclear Regulatory Commission Region II Regional Administrator. The annual report shall contalu as a minimum, the following information:

a.

A brief narrative summary:

i.

Operating experience including a cross section of experiments performed.

1 ii.

Changes in performance characteristics related to reactor safety that occurred during the reporting period iii.

Results of surveillance, tests and inspections.

b.

Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality.

c.

The number of emergency shutdowns and inadvertent SCRAMS, including reasons therefore.

Amendment 11 46 December 15, 1989

Appendix A Technical Specifications d.

Discussion of the major corrective maintenance operations performed during the period, including the effect, if any, on the safety of operation of the reactor, A brief description, including a summary of the safety evaluations of e.

changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR.

f.

A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge, including:

Liquid Waste (summarized by quarter) 1.

Radioactivity released during the reporting period:

(1)

Number of batch releases.

(2)

Total radioactivity released (in microcuries).

_(3)

Total liquid volume released (in liters).

(4)

Diluent volume required (in liters).

(5)

Tritium activity released (in microcuries).

(6)

Total (yearly) tritium released.

(7)

Total (yearly) activity released.

ii.

Identification of fission and activation products:

Whenever the concentration of radioactivity in the waste tank at the time of release without dilution exceeds 4 x 10~5 Ci/ml, as determined by a gross beta gamma count, the dried residue of a one liter sample shall be analyzed prior to release for principal gamma emitting radionuclides. An estimate of the quantities present shall be reported for each of the identified nuclides.

iii.

Disposition of liquid effluents not releasable to the sanitary sewer system:

Any waste tank containing liquid effluent failing to. meet the requirements of 10 CFR 20, Appendix B, reported hereunder, to include the following data:

(1)

Method of disposal.

Amendment 11 47 December 15, 1989

Appendix A Technical Specincations (2)

Total radioactivity in the tank (in anrocuries) prior to disposal.

(3)

Total volume of liquid in tank (in lian).

i (4)-

The dried residue of a one liter sample shall be analyzed for the principal gamma-emitting rad %cacudes. The 4

identified isotope composition with eni.cr.ated concentrations shall be reported. The tritium contec.t shall be included.

Ganeous Waste (summarir$d on a monthly basis)

[

i.

Radioactivity discharged during the repordog period'(in curies) for:

(1)

Gases

'(2)

Particulates, with half lives greater than eight days.

t li.

The MPC used and the estimated activity (in curies) discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis.

Solid Waste L

i.

The total amount of solid waste packaged (in cubic feet).

~

ii.

The total activity involved (in curies).

iii..

The dates of shipme,nt 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 facility.

i.

A description of environmental surveys performed outside the facility.

Amendment 11

• 48 December 15, 1989

Appendix 4 Technical Specincations L

6.8 Retention of Records Records and logs of the following items, as a minimum, shall be kept in a manner convenient for review and shall be retained as detailed below. In addition, any additional federal requirement in regards to record retention shall be met.

Records to 5 retained for a period of at least five (5) years:

a.

1.

Normal plant operation and maintenance.

ii.

Principal maintenance activities, iii.

Reportable events.

iv.

Equipment and components surveillance activities.

v.

Experiments performed with the reactor, b.

Records to be retained for the life of the facility:

1.

Gaseous and liquid radioactive waste released to the environs.

ii.

Results of off site environmental monitoring surveys, iii.

Radiation exposures for all PULSTAR personnel, iv.

Results of facility radiation and contamination surveys.

v.

Fuel inventories and transfers, vi.

Drawings of the reactor facility.

c.

Records to be retained for at least one training cycle:

i.

Records of retraining and requalification of certified operating personnel shall be maintained at all times the individual is employed, or until the certification is renewed.

Amendment 11 49 December 15,1989

POWER - FLOW Safety Limit. Curve 5

C Pool Depth Safety Limit.- 14' min E

Pool Temp Safety Limit '- 120 F max "4

u 1007. Flow = 3.8 MW o

ct

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Percent of Full 500 gpm Core Flow FIGURE 2.1-1 l

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m ii ii i ii i i i NORTH --CAROLINA 4 STATE ! UNIVERSITY PULSTAR REACTOR ORGANIZATIONAL CHART CHANCELLOR NORTH CAROUNA STATE UNfvERSffY' l.

RADUTION PROTECTM COL OF COUNCIL ~

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HEAD DEPARTMENT OF j

REACTOR SAFEGUAROS NUCLEAR ENGINEERING

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DIRECTOR NUCLEAR REACTOR PROGRAM l

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l REACTOR HEALTH ASSOCIATE D! RECTOR l

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PHYSICIST NUCLEAR REACTOR PROGRAM ADVISE & UAISON i'

REACTOR OPERATIONS l -

(SENIOR REACTOR OPERATOR)

MANAGER I

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CHIEF REACTOR OPERATOR CNIEF OF REACTOR (REACTOR OPERATOR)

MAINTENANCE SAR FIGURE Ii-I TECHNICAL SPECIFICATIONS FIGURE 6-1 PULSTAR OPERATIONS MANUAL FIGURE 2-1

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