ML111950372

Response to NRC Request for Additional Information (RAI) Questions 1 Through 37 (Non-Financial), Technical Specifications. Redacted Version

ML111950372
Person / Time
Site:	05000128
Issue date:	05/27/2011
From:	Reece W Texas A&M Univ
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Lynch, Steven - NRR/DPR/PRPB, 415-1524
References
2011-0025
Download: ML111950372 (109)

Text

Texas A&M University Nuclear Science Center License No. R-83 Docket No. 05000128 Response to NRC Request for Additional Information (RAI)

Questions 1 through 37 (Non-Financial),

Technical Specifications Redacted Version*

Security-Related Information Removed

• Redacted text and figures blacked out or denoted by brackets

TEXAS ENGINEERING EXPERIMENT STATION TEXAS A&M UNIVERSITY 3575 TAMU COLLEGE STATION, TEXAS 77843-3575 NUCLEAR SCIENCE CENTER 979/845-7551 May 27, 2011 FAX 979/862-2667 Document Control Desk U.S. Nuclear Regulatory Commission 2011-0025 Washington, DC 20555-0001

Subject:

Response to NRC Requests for Additional Information Questions I through 37 (Non-Financial) for the Nuclear Science Center Reactor (NSCR, License No. R-83, Docket 50-128)

To Whom It May Concern:

The Texas A&M University System, Texas Engineering Experiment Station (TEES), Nuclear Science Center (NSC, License No. R-83) operates a LEU, IMW, TRIGA reactor under timely renewal. In December, 2003 the NSC submitted a Safety Analysis Report (SAR) as part of the license renewal process. In December, 2005 a conversion SAR (Chapter 18) was submitted resulting in an order to convert from the NRC. In July 2009, the NSC submitted an updated SAR, dated June 2009, to the Nuclear Regulatory Commission (NRC). This updated 2009 version of our SAR incorporated the information from the conversion SAR and the startup of the new LEU reactor core. On November 13, 2009 the NRC submitted a Request for Additional Information as a part of the review process. This request included 37 questions about the technical aspects of the NSC's SAR submittal. Attached to this letter are the NSC's response to Questions 1-37 of the NRC's non-financial RAls. Please note in the attachment that text from the NRC's letter has been italicized and the NSC's non-italicized response immediately follows each question.

If you have any questions, please contact Jim Remlinger, Jerry Newhouse, or W. Dan Reece at 979-845-7551.

These RAIs and the enclosed Technical Specifications have been reviewed by our Reactor Safety Board and they have requested 30 days to review these documents for errors, but approved submittal to the NRC.

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 27, 2011.

W. D. Reece Director, Nuclear Science Center Xc: 2.1 1/Central File Duane Hardesty, NRC Project Manager RESEARCH AND DEVELOPMENT FOR MANKIND http://lnsc.tamu.edu

RAI 1

The attached report "Evaluation of Pool Temperature on Fuel Temperature and MDNBR for the NSC TRIGA Reactor" concludes that the increase of inlet pool temperature to 60 C does not cause the fuel centerline on the hottest fuel element to exceed the LSSS and the DNBR with the most conservative calculations is 1.54. The NSC further concludes there is no need for a TS or alarm for coolant temperature and the fuel is well protected by our LSSS setting with its alarm and scram. The NSC operators are acutely aware of coolant temperature since they record pool temperature every 30 minutes. The NSC has a large pool so that pool temperature increases gradually. This feature provides ample time for reactor operators to react to increasing pool temperatures obviating the need for an automatic scram. We will create an LCO for pool temperature and add to our SOPs that the director, or his designee, will be notified of pool temperatures over 50 C and that reactor operation is not allowed for pool temperatures above 60 C.

14.3.8.3 PrimaryCoolant Temperature Applicability This specification applies to the maximum allowable primary coolant temperature.

Objective The objective is to maintain fuel temperature less than the LSSS, to maintain the departure of nucleate boiling ratio (DNBR) greater than unity, and to limit any degradation of the reactor systems.

Specification

1) The Director, or his designee, shall be notified when pool temperature reaches 50 C, with further operation dependent on his approval.

2) The reactor shall not be operated when pool temperature exceeds 60 C.

Basis Calculations show that reactor operations with a pool temperature of 60 C will not risk reaching a fuel temperature greater than the LSSS, or a DNBR of unity.

Surveillance Requirement added to 14.4.3.2 Reactor Safety Systems. An increase in pool temperature from 30C to 60C can marginally increase the corrosion rates for the stainless steel pool liner and on the aluminum components in the pool. However, these corrosion rates are trivial compared to the cladding on the fuel at and the aluminum top handles. The fuel cladding and top handles were used for 25 years in our previous HEU core with no observable damage. There is a possibility of softening the exchange resins and making demineralization less efficient, but this is covered by the corrosion and pH TS limits.

The pulses happen so quickly that the governing equations are derived assuming adiabatic conditions during the pulse. If we pulse at 60C then we need to limit our reactivity insertion so that we do not exceed our TS limit of 830C. By procedure we collect our fuel temperature before pulsing and include its effect on the pulse. We'll move the equation out of the TS basis into the SAR.

RAI 2

Work at Argonne National Laboratory on full and partial LOCA scenarios was published in 2009 (Analyses of Loss-of-Coolant Accidents for TRIGA-Fueled Research Reactors at Washington State University, the University of Wisconsin, and Oregon State University, F.E. Dunn, E.E. Feldman, T. Sofu, and J.E. Matos, RERTR Program, Nuclear Engineering Division, Argonne National Laboratory 60439).

This report is available at no cost, at http://www.osti.gov/bridge. It is also available on paper to the U.S.

Department of Energy and its contractors, for a processing fee, from: U.S. Department of Energy, Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831-0062 (reports@adonis.osti.gov).

In this report, the Washington State, University of Wisconsin, and Oregon State reactors were analyzed for partial LOCA events. The important parameters as described in this report are the power production in the hottest rod and the amount of fuel meat left covered by the failure of a beam port. For the University of Wisconsin reactor, Argonne found:

"The peak rod power was calculated to be 19.3 kW at this reactor power level. Since the uncertainty in the measured reactor power level is 2%, this rod power is increased by 2% to 19.7 kW. For a decay heat level corresponding to continuous full-power operation for 7 days/week and a drain time of 15 minutes, a fuel rod power level of 24.0 kW is calculated to result in a peak fuel surface temperature of 700'C. For a rod power of 19.7 kW, the peak fuel surface temperature would be 585°C (see Appendix A, Table A6, linear fit).

For a decay heat curve based on continuous operation at 1.0 MW for 5 days/week for 40 years and a drain time of 15 minutes, Table 2 shows that a peak fuel surface temperature of 700'C would be reached for a fuel rod power of 27.2 kW, which is much higher than the calculated peak rod power of 19.7 kW, including a 2% uncertainty. The peak fuel surface temperature was calculated to be 525°C under these conditions (see Appendix A, Table A6, linear fit)." (Argonne, op. cit., p. 10)

They further analyze for a partial LOCA.

"The centerlines of the four UWNR beam tubes are aligned with the core mid-plane, which is located at 7.5 inches above the bottom of the fuel. Since the beam tubes are 6 inches in diameter, the lowest initial water level for the partial LOCA analysis is 4.5 inches above the bottom of the fuel. For a fuel rod power of 19.7 kW and a drain time of 15 minutes, the analysis in Appendix B predicted a peak fuel temperature of 578°C. Since this temperature is 1220 C below a maximum fuel temperature in air of 700'C, a LOCA initiated by a failure of one of the UWNR beam ports will not result in failure of the hottest fuel rod."

(Argonne, op. cit., p. 11)

The hottest rod in the NSC reactor is calculated to be 17.4 kW. With a 2% uncertainty, it would be 17.8 kW. Clearly we are below the power used for the Wisconsin analysis and well below the threshold for fuel damage during partial LOCAs according to the Argonne analysis. The Argonne analysis showed that 4.5 inches of fuel meat would be left underwater at the Wisconsin reactor during a partial LOCA. The length of fuel meat left covered by water if Beam Port 1 or 6 is sheared is the same as the analysis of the Wisconsin reactor. Beam Ports I and 6 centerlines are at the center of the fuel and the limiting part of the beam ports is a 6.00 inch inner diameter pipe. Since the fuel meat is 15 inches long, the part under water is then: 15/2 - 6/3 = 4.5 inches. Therefore, for our Beam Ports I and 6, we are bounded by the Argonne calculations.

We also have a thermal column as shown in the attached Figure 1. First, it is a most improbable occurrence that the thermal column could be compromised - large lifts are not done over the stall where

the column is located. Nevertheless, if the coupler were sheared, an aluminum plate between the coupler and the thermal column extension still separates the pool from the column. If the extension were sheared, then the pool is drained to 38.75 inches above the pool floor which is below the grid plate, so the full LOCA analysis applies. This full LOCA analysis has already shown that the fuel cladding is not breached for the hottest rod.

RAI 3

As stated in 10CFR50, in part:

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

The design basis accident (DBA) performed in section 13 assumes the loss of integrity of the fuel cladding for one fuel element and the simultaneous loss of pool water resulting in fission product release.

The calculations were based on the fission product release.

The accumulated dose to staff following the onset of accident is calculated to be 0.25 mrem for the 5 minute duration after release. This is in agreement with the I OCFR20.1201 requirements (TEDE of 5 rem).

The maximum exposure to personnel in unrestricted area calculations includes the dose to unshielded gamma rays and ground shine from the plume passage. The whole body dose (TEDE) calculated for the member of the public is 6.2 mrem and it satisfies the 10CFR20.1301 requirements (TEDE of 100 mrem).

MCNP calculations were performed modeling the containment building to calculate the dose at the fence line to members of the general public for dose arising from a completely uncovered (but intact) core. In all cases studied, the maximum dose to a person at the fence line is belowl.0 mrem/hr.

The dose in the control room, arising principally from scattered radiation, from the uncovered core is calculated to be 4.27 rem/hr. With a five-minute exit time this would lead to a dose of 356 mrem, well below the 10CFR20 requirements. In actually, the quickest the pool could drain is on the order of and even 3 feet of water cuts the dose by an order of magnitude or more.

RAI 4

NSC Response: The NSC considers some details of the pool level alarm system to be Safeguards Information, but the following information is provided.

a. A robust system is in place that identifies loss of pool level. An alarm is received both locally and at a 24- hour manned facility. Procedures are in place for a rapid response from both University Police Department and NSCR personnel when an alarm level is reached. A "call out" list for NSCR responders is provided to appropriate response entities.

b. Please note that TS 14.5.7 has the following specification: "4. A pool level alarm shall indicate loss of primary coolant before or equal to the pool level dropping to 10% below the normal operating level."

c. All leakage will end up in a waste water transfer sump. This water ends up in our waste storage tanks.

Water in these tanks are analyzed for radioactive contamination prior to their release to a waste water processing facility. The NSC has detailed operating procedures for response to a pool water leak. Based on years of waste water tank releases the pool radioactive material content is normally less than 10% of the limit for release to sewerage as listed in 10 CFR Part 20, Appendix B.

d. The 17 minute time is simply a conservative point of reference. As the TS states, it is the time to uncover the core when a single 10" line on the pool floor is sheared. The calculation assumes the pipe is sheared in the sump with no makeup or friction losses during drainage.

e. Again, a low level alarm identifies a loss of pool water. During operation change in pool level is easily noted since we have a large open pool and our control room is on the same level and adjacent to the pool.

As indicated in "a" above the alarm is sent to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dispatch who notifies appropriate responders.

RAI 5

The Nuclear Science Center will modify the TS 14.1 to include the terms above using the American National Standards Institute (ANSI), American Nuclear Society (ANS) 15.1. ANSI/ANS 15.1-2007, Section 1.3.

14.1.7 Core Configuration The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and regulating/control/transient rods occupying the core grid.

14.1.9 Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff = 1) at reference core conditions or at a specified set of conditions.

14.1.16 License The written authorization, by the responsible authority, for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or facility requiring licensing.

14.1.17 Licensee An individual or organization holding a license.

14.1.28 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.

14.1.34 Reactor Operator An individual who is licensed to manipulate the controls of a reactor.

14.1.49 Senior Reactor Operator An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.

RAI 6

The NSC proposes eliminating the term protective devices in 14.1.19, redefining the Limiting Safety System Setting while adding definitions for protective actions and operator.

14.1.19 Limiting Safety System Setting The limiting safety system setting is the setting for initiating automatic protective action.

14.1.27 Operator See Reactor Operator.

14.1.28 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.

RAI 7

The NSC will eliminate the use of the term "Cold Critical," and replace it with the term "Reference Core Condition," and add the ANSI standard definition for "Reference Core Condition" to the definitions section.

14.1.38 Reference Core Condition The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (<0.30 dollar).

14.3.1.5 Maximum Excess Reactivity Applicability This specification applies to the maximum excess reactivity, above reference core condition, which may be loaded into the reactor core at any time.

14.3.6.1 Reactivity Limits Basis

2. The maximum worth of a single experiment is limited so that its removal from the reference core condition reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

RAI 8

The NSC will modify our definition for weekly to follow ANSI/ANS 15. 1. ANSI/ANS- 15.1-2007.

Additionally, the definition for Biennially was changed to match the standards.

14.1.55 Weekly Items which are designated by these specifications or by standard operating procedures to be performed weekly shall be performed at intervals not to exceed 10 days.

14.1.4 Biennially Items which are designated by these specifications or by standard operating procedures to be performed biennially should be performed at least once every 24 months, but at no time shall the time period between consecutive performances of these items exceed 30 months.

RAI 9

The Nuclear Science Center will modify the TS 14.1.31(3) [now TS 14.1.39(3)] to remove the stipulation and conform to the American National Standards Institute (ANSI), American Nuclear Society (ANS) 15.1. ANSI/ANS-15.1.-2007, Section 6.7.2.

14.1.39 Reportable Occurrence A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required. (Note: Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable provided that the minimum numbers of components or systems specified or required perform their intended reactor safety function.)

RAI 10

The NSC will change the Tech Spec to read:

14.3.1.1 Steady State Operation Applicability This specification applies to the energy generated in the reactor during steady state operation.

Obiective The objective is to ensure that the fuel temperature safety limit will not be exceeded during steady state operation.

Specifications The reactor power level shall not exceed 1.0 megawatt (MW) during steady state operation.

Basis Calculations show that reactor operations with a pool temperature of 60 C will not risk reaching a fuel temperature greater than the LSSS, or a DNBR of unity for 1MW steady state.

RAI 11 TAMU TS 14.3.2 does not define an LCO for minimum number of operable control rods. We will modify the TS 14.3.2.3 Scram Time to include this LCO.

14.3.2.3 Minimum Number of Operable Scrammable Control Rods Applicability This specification applies to the minimum number of operable scrammable control rods in the core, where operable is specified in terms of maximum scram time from the instant that the fuel temperature safety channel variable reaches the Limiting Safety System Setting.

Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Specification During standard operation all scrammable control rods shall be operable. The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable rod reaches its fully inserted position shall not exceed 1.2 seconds. During core manipulations, i.e. core loading and unloading, all installed scrammable control rods shall be operable.

Basis This specification ensures that the reactor will be promptly shutdown when a scram signal is initiated.

Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.

RAI 12

The NSC proposes to change the technical specification wording to show that the Fuel Element Temperature and High Power Level safety channels will "Scram at the LSSS" and "Scram < 125%."

RAI 13

TS 14.3.2.2 will be updated to read:

1.1.1.1 14.3.2.2 Reactor Safety Systems and Interlocks Applicability This specification applies to the reactor safety system circuits and interlocks.

Objective The objective is to specify the minimum number of reactor safety system channels and interlocks that must be operable for safe operation.

Specifications The reactor shall not be operated unless the safety circuits and interlocks described in Table 14-2 are operable, however during periods of maintenance, surveillance, calibration and repair, the reactor may be operated as necessar, to calibrate and ensure channel operability provided that all the conditions of the TS are met during the operation.

Table Error! No text of specified style in document.-l: Safety Circuits and Interlocks Required for Operation

., . Operating 0 Mode,

" Number

"..S afety C Safey

,.':":"' ha n*hnne I  !  :.:i !, .:: :. . .. ...Operable

.Function

.. . ..: S.S. .. ......

Pulse Fuel Element Temperature I Scram < the LSSS X X High Power Level 2 Scram < 125% X -

Console Scram Button I Scram X X High Power Level Detector Power 2 Scram on loss of supply voltage X Supply Preset Timer I Transient Rod Scram 15 seconds or -

less after pulse Prevent withdrawal of Shim Safeties Log Power I at <4x I0-3 W X Log Power I Prevent pulsing above I kW X Transient Rod Position I Prevent application of air unless fully X inserted

Shim Safety and Regulating Rod Position I Prevent withdrawal X Pulse Stop Electro-Mechanical Interlock I Prevent application of air unless x pulse stop is installed Basis

1. The fuel temperature and high power level scrams provide protection to assure that the reactor can be shutdown before the safety limit on fuel element temperature will be exceeded.

2. The manual console scram allows the operator to shut down the system if an unsafe or abnormal condition occurs.

3. In the event of failure of the power supply for a high power level safety detector, operation of the reactor without adequate instrumentation is prevented.

4. The preset timer ensures that the reactor power level will reduce to a low level after pulsing.

5. The interlock to prevent startup of the reactor at power levels less than 4 x 10-3 W, which corresponds to approximately 2 cps, assures that sufficient neutrons are available for proper startup.

6. The interlock to prevent pulsing at powers above 1 kW assures that the magnitude of the pulse will not cause the fuel element temperature safety limits to be exceeded.

7. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing of the reactor in steady state mode.

8. The interlock to prevent the withdrawal of the shim safeties or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive period.

9. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted prevents a reactor pulse of sufficient worth to exceed the temperature safety limit.

RAI 14

A) (1) We are natural convection cooled reactor. We have no pump requirements. There is no need for a limiting condition of operation requirement.

(2) We have covers for the primary intake and outflow. This is covered in TAMU TS 14.5.7.

(3) The reactor shall not be operated if the pool level is below the intake for the diffuser system.

(4) No LCO is required. A daily-checked easily visible gauge on the pool wall is satisfactory visual indication of water loss and is used to determine makeup. Daily readings of pressure of the primary and secondary cooling systems ensure non-leakage through the heat exchanger.

(5) This is already in our Technical Specifications. TAMW TS 14.5.4 and new TAMU TS 14.4.4.1 (6) The NSC operates a large surface area non-enclosed pool in a facility with at least 180,000 cubic feet of free space, so the concentration of hydrogen is negligible. This standard is not applicable.

(7) Our design basis accident does not include any emergency cooling system. We have no limiting condition of operation requirement.

(8) By Tech Spec limits, the reactor pool quality, in terms of both chemistry and radioactive contaminants, will always be such that it could be discharged directly to the sanitary sewer. We do not operate the reactor if the primary system (pool) contains appreciable anions that will activate so to control the spread of contamination. Given this, the chances of contaminating the secondary system are negligible. We do not need a limiting condition of operation requirement.

(9) This is already in our Technical Specifications. TAMU TS 14.3.8 The following TS will be added:

14.3.8.2 Primary Coolant Level and Leak Detection Applicability This specification applies to the physical level of primary coolant that must be in the pool and requirements for leak detection for reactor operation.

Objective The objective is to ensure proper shielding and cooling of the reactor and the ability to detect leaks.

Specification

1) The reactor shall not be operated if the pool level is lower than 18 feet above the top of the core.

2) The reactor shall not be operated if the pool level alarm is not operable, and capable of producing an alarm at a constantly monitored off-site facility.

Basis

1) The intake to the diffuser system is approximately 18 feet above the core. Setting this level as the specification will both ensure the availability of the diffuser system, and provide more than adequate shielding and cooling for the reactor.

2) An operable pool level alarm that notifies off-site will ensure proper notification if a low pool level is reached.

14.4.3.2 Reactor Safety Systems Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of the control and safety systems.

Objective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.

Specifications

1. A channel check of each of the reactor safety system channels for the intended mode of operation shall be performed before each day's operation or before each operation extending more than one day, except for the pool level alarm channel, which shall be tested weekly.

2. Whenever a reactor scram caused by high power level or high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety limit was exceeded.

3. A calibration of the fuel element temperature measuring channels shall be performed semiannually.

4. A calibration of the pool temperature measuring channel shall be performed semiannually.

5. A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated.

6. Reactor safety systems which include interlocks shall be tested at least annually.

Basis

Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. If the period between operations extends beyond a year, then the annual channel test requirement will assure operability.

RAI 15

The technical specifications 14.3.3.1 and 14.3.3.2 include an extended note discussing a one hour limit and a basis to reflect the change in specifications 14.3.3.1 (1) and 14.3.3.2 (1).

14.3.3.1 Operations that Require Confinement Applicability This specification applies to confinement requirements during operation of the reactor and the handling of radioactive materials.

Objective To maintain normal or emergency airflow into and out of the reactor building during operations that produce or could potentially produce airborne radioactivity.

Specification Confinement of the reactor building shall be required during the following operations:

1. Reactor operating Note: For periods of maintenance to the central exhaust fan, entry doors to the reactor building shall remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Handling of radioactive materials with the potential for airborne release.

Basis

1. This 41 basis applies during the conduct of those activities defined as reactor operations.

is produced during operation of the reactor in experimental facilities and in the "Ar reactor pool; thus, air control within the building and the exhaust system in necessary to maintain proper airborne radiation levels in the reactor building and release levels in the exhaust stack. Other radioactivity releases to the reactor building must be considered during reactor operation, such as fission product release from a leaking fuel element or a release from fixed experiments in or near the core. Shutdown of the central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The handling of radioactive materials can result in the accidental or controlled release of airborne radioactivity to the reactor building environment or direct release to the building exhaust system. In these cases, the control of air into and out of the reactor building is necessary.

14.3.3.2 Equipment to Achieve Confinement Applicability This specification applies to the equipment and controls needed to provide confinement of the reactor building.

Objective The objective is to ensure that a minimum of equipment is in operation to achieve confinement as specified in Section 14.3.3.1 and that the control panel for this equipment is available for normal and emergency situations.

Specifications

1. The minimum equipment required to be in operation to achieve confinement of the reactor building shall be the central exhaust fan.

Note: During periods of maintenance to the central exhaust fan, entry doors to the reactor building shall remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Controls for establishing the operation of the ventilation system during normal and emergency conditions shall be available in the reception room.

Basis

1. Operation of the central exhaust fan will achieve confinement of the reactor building during normal and emergency conditions when the controls for air input are set such that the central exhaust fan capacity remains greater than the amount of air being delivered to the reactor building. The exhaust fan has sufficient capacity to handle extra air intake to the building during momentary opening of doors. Shutdown of the central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The control panel for the ventilation system provides for manual selection of air input to the reactor building and the automatic or manual selection of air removal. The air supply and exhaust systems work together to maintain a small negative pressure in the reactor building. These controls are available in the reception room for accessibility during emergency conditions.

RAI 16

The ventilation system includes the central exhaust and air handlers. While the air handlers are not required for operation, the louvers that isolate the intakes are required to close when the air handler is shut down. This is checked weekly during the ventilation system checks. The technical specification will now explicitly state this difference and the conditions that follow.

14.3.4 Ventilation System Applicability This specification applies to the operation of the facility ventilation system.

Obiective The objective is to ensure that the ventilation system can mitigate the consequences of the possible release of radioactive materials resulting from reactor operation and radioactive material handling.

Specification The reactor shall not be operated unless the following parts of the facility ventilation system are operable:

1) Central Exhaust Fan

2) Intake louvers on the air handlers The reactor may be operated during periods of maintenance up to one hour. During this maintenance period the reactor building doors shall remain closed except for momentary opening for personnel entry or exit.

The ventilation system shall be secured automatically by signals from the stack particulate, fission product, or stack gas (xenon) facility air monitor.

Basis During normal operation of the ventilation system, the concentration of 41 Ar in unrestricted areas is below the effluent concentration (SAR Chapter 11). In the event of a substantial release of airborne radioactivity, the ventilation system will be secured automatically. Therefore, operation of the reactor with the ventilation system shutdown for short periods of time to make repairs ensures the same degree of control of release of radioactive materials. Moreover, facility air monitors within the building independent of those in the ventilation system will give warning of highlevels of radiation that might occur during operation with the ventilation system secured.

The ventilation system must be shutdown for weekly operability check. The one hour limit provides enough time to complete these checks.

RAI 17

The NSC proposes to make changes to applicable technical specification(s) to include the following:

Note: When required monitors are inoperable, or for periods of maintenance, the intent of this specification will be satisfied if this monitor is replaced for a period of no more than 1 week with a portable gamma sensitive instrument having its own alarm which shall be kept under visual observation.

If two of the above monitors are not operating, the reactor will be shutdown.

RAI 18

TS 14.3.6.2 Material Limitations will be updated to read:

14.3.6.2 MaterialLimitations Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

Obiective The objective is to prevent damage to the reactor or excessive release of radioactivity by limiting materials quantity and radioactive material inventory of the experiment.

Specifications I. Explosive materials in quantities greater than 5 pounds (TNT) shall not be allowed within the reactor building. Irradiation of explosive materials shall be restricted as follows:

a. Explosive materials in quantities greater than 25 milligrams (TNT) shall not be irradiated in the reactor pool. Explosive materials in quantities less than 25 milligrams (TNT) may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container.

b. Explosive materials in quantities greater than 25 milligrams (TNT) shall be restricted from the reactor pool, the upper research level, the demineralizer room, cooling equipment room and the interior of the pool containment structure.

c. Explosive materials in quantities greater than 5 pounds (TNT) shall not be irradiated in experimental facilities.

d. Cumulative exposures for explosive materials in quantities greater than 25 milligrams (TNT) shall not exceed 1012 n/cm 2 for neutron or 25 Roentgen for gamma exposures.

2. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 10 Ci.

3. Significant amounts of corrosive materials used in a reactor experiment shall be double encapsulated. Exceptions may only be made if a detailed analysis and/or prototype testing with small amounts of materials demonstrates that the experiment presents negligible risk.

Basis

1. This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving explosive materials.

a. This specification is intended to prevent damage to the reactor core and safety related reactor components located within the reactor pool in the event of failure of an experiment involving the irradiation of explosive materials. Limited quantities of less than 25 milligrams (TNT) and proper containment of such experiment provide the required safety for in-pool irradiation provided that the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container.

b. This specification is intended to prevent damage to vital equipment by restricting the quantity and location of explosive materials within the reactor building.

Explosives in quantities exceeding 25 milligrams (TNT) are restricted from areas containing the reactor bridge, reactor console, pool water coolant and purification systems, and reactor safety related equipment.

c. The failure of an experiment involving the irradiation of up to 5 pounds (TNT) of explosive material in an experimental facility located external to the reactor pool structure will not result in damage to the reactor or the reactor pool containment structure.

d. This specification is intended to prevent any increase in the sensitivity of explosive materials due to radiation damage during exposures.

2. The 10 Ci limitation on Iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR 20 for an unrestricted area.

3. This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving corrosive materials.

The analysis supporting this TS was performed and approved by the NRC in an NSC license amendment in 1979.

RAI 19

We will modify TAMU TS 14.3.6.2 to address corrosive materials.

14.3.6.2 Material Limitations Specifications

3. Significant amounts of corrosive materials used in a reactor experiment shall be double encapsulated.

Exceptions may only be made if a detailed analysis and/or prototype testing with small amounts of materials demonstrates that the experiment presents negligible risk.

Basis 3.This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving corrosive materials.

RAI 20

We will modify TAMU TS 14.3.6.1 to address the sum of the absolute value of all experiments in the reactor.

1.1.1.2 14.3.6.1 Reactivity Limits Applicability This specification applies to the reactivity limits on experiments installed in the reactor and its experimental facilities.

Obiective The objective is to ensure control of the reactor during the handling of experiments adjacent to or in the reactor core.

Specifications The reactor shall not be operated unless the following conditions governing experiments exist.

1. The reactivity worth of any single, movable or unsecured experiment shall be less than one dollar.

2. The reactivity worth of any secured experiment shall be less than two dollars.

3. The sum of the absolute reactivity of all experiments shall be less than five dollars.

Basis

1. This specification is intended to ensure that the worth of a single unsecured experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were suddenly inserted. This does not restrict the number of unsecured experiments adjacent to or in the reactor core except by worth.

2. The maximum worth of a single experiment is limited so that its removal from the reference core condition reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be secured, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

3. This limit poses a restriction on the number of experiments being run at any given time to prevent excessive positive and negative reactivity loading.

RAI 21

The NSC will change Basis 2 to read:

Basis

2. In the NRC's safety evaluation supporting amendment 15 to our facility license the commission concluded that the NSC's request for a total inventory limit of Xe-125 of 3500 Ci was reasonable based on the facility experimental design and operating plan for Iodine Production.

RAI 22

TS 14.3.8.1 Primary Coolant Purity will be updated to read:

14.3.8.1 PrimaryCoolant Purity Applicability This specification applies to the quality of the primary coolant in contact with the fuel cladding.

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be no higher than 5 x 10-6 mhos/cm (5 iiSiemens/cm).

2. The pH of the bulk pool water shall be between 5.5 and 7.5.

3. The concentrations of radionuclides in the bulk pool water shall be no higher than the values presented for water in 10 CFR Appendix B to Part 20 Table 2.

Basis A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations.

RAI 23

The following will be added to TAMU TS 14.4 to address surveillance deferment during shutdown:

14.4.8 Surveillance Deferment Applicability This specification describes which surveillance requirements may be deferred during reactor shutdown, and which shall be performed prior to resuming reactor operations.

Obiective The objective is to give facility personnel the ability to avoid conducting unnecessary surveillances during periods of reactor shutdown.

Specifications Possible to Defer TS During Shutdowns? Required prior to operations?

1. 14.4.2.1 Steady State Operation Yes Yes

2. 14.4.2.2 Pulse Mode Operation Yes Yes

3. 14.4.2.3 Shutdown Margin Yes Yes

4. 14.4.2.4 Excess Reactivity Yes Yes

5. 14.4.2.5 Reactor Fuel Elements No N/A

6. 14.4.3.1 Reactor Control Systems No N/A

7. 14.4.3.2 Reactor Safety Systems Yes, except pool level Yes

8. 14.4.3.3 Scram Time Yes Yes

9. 14.4.4.1 Coolant Radioactivity Analysis Yes - Quarterly Yes

10. 14.4.4.2 Coolant Conductivity and pH Yes - Quarterly Yes

11. 14.4.5 Equipment to Achieve Confinement:

Ventilation System Yes Yes

12. 14.4.6 Radiation Monitoring Systems and Effluents No N/A

13. 14.4.7 Experiments Yes Yes

Basis

1. It is impossible to perform a calorimetric while the reactor is shutdown; a calorimetric shall be performed prior to resuming operations.

2. It is impossible to perform pulse surveillance while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

3. It is impossible to determine control rod worth while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

4. It is impossible to determine control rod worth while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

5. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

6. The possibility for control rod degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

7. With the exception of the pool level alarm, whose surveillance shall not be deferred, it is not necessary to conduct surveillance on reactor safety systems, which are effectively automatic scram systems, while the reactor is shutdown. The surveillance shall be performed prior to resuming operations.

8. It is impossible to determine scram time while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

9. The possibility for cladding degradation or damage still exists while the reactor is shutdown; however it is much less likely. Therefore the periodicity of this surveillance may be changed to quarterly. The surveillance shall be performed prior to resuming operations.

10. Fuel cladding degradation is still undesirable while the reactor is shutdown; however chemically contaminating the pool is much less likely. Therefore the periodicity of this surveillance may be changed to quarterly. The surveillance shall be performed prior to resuming operations.

11. The ventilation system is only required during reactor operations or radioactive materials handling. The surveillance shall be performed prior to resuming operations or handling.

12. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

13. No experiments can be installed in the reactor while it is shutdown; the surveillance shall be performed prior to resuming operations.

Additionally, TS 14.3.8.2 has been added defining an LCO for pool level:

14.3.8.2 Primary Coolant Level and Leak Detection Applicability This specification applies to the physical level of primary coolant that must be in the pool and requirements for leak detection for reactor operation.

Objective The objective is to ensure proper shielding and cooling of the reactor and the ability to detect leaks.

Specification

1) The reactor shall not be operated if the pool level is lower than 18 feet above the top of the core.

2) The reactor shall not be operated if the pool level alarm is not operable, and capable of producing an alarm at a constantly monitored off-site facility.

Basis

1) The intake to the diffuser system is approximately 18 feet above the core. Setting this level as the specification will both ensure the availability of the diffuser system, and provide more than adequate shielding and cooling for the reactor.

2) An operable pool level alarm that notifies off-site will ensure proper notification if a low pool level is reached.

RAI 24

The excess reactivity is maximum at initial core loading and decreases monotonically for 30/20 LEU fuel.

(See SAR.) However we agree that it would be helpful to document excess reactivity at least biennially and after any significant core or control rod change. We will add the following Technical Specification:

14.4.2.4 Excess Reactivity Applicability This specification applies to the surveillance requirements of reactor excess reactivity.

Objective The objective is to verify that requirements on excess reactivity are met for operational cores.

Specification The excess reactivity shall be determined biennially and following significant core configuration and/or control rod changes.

Basis The excess reactivity of the core is measured to ensure that during all states of operation criticality can be maintained for licensed operational limits. With the accumulation of fission product poison buildup and fissile material burnup, excess reactivity must be available for power transients and maintaining criticality.

RAI 25

The TS will be changed as follows:

14.4.2.3 Shutdown Margin Applicability This specification applies to the surveillance requirement of control rod calibrations and shutdown margin.

Objective The objective is to verify that the requirements for shutdown margins are met for operational cores.

Specification The reactivity worth of each control rod and the shutdown margin shall be determined annually and following significant core configuration and/or control rod changes.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to insure no significant changes in the shutdown margin.

RAI 26

The TS will be changed as follows:

14.4.3.1 Reactor Control System Applicability These specifications apply to the surveillance requirements for reactor control systems.

Obiective The objective is to verify the condition and operability of system components affecting safe and proper control of the reactor.

Specifications

1. The control rods shall be visually inspected for deterioration biennially.

2. Operability tests of the control rod mechanism shall follow modification or repairs.

3. The Transient Rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated semiannually.

Basis

1. The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation of the reactor.

2. These tests provide verification that the control rod has full travel and that the rod drop time is within specification.

3. Inspection and maintenance of the transient rod drive assembly reduces the probability of failure of the system due to moisture-induced corrosion of the pulse cylinder and piston rod assembly.

RAI 27

The NSC will add a point to the specifications of 14.4.3.2 Reactor Safety Systems to specifically address reactor control interlocks.

5. Reactor safety systems which include interlocks shall be tested at least annually.

The basis will also be changed to read:

Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. If the period between operations extends beyond a year, then the annual channel test requirement will assure operability.

Additionally, Specification No. 4 will be rewritten to add the word channel for clarity.

4. A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature.

RAI 28

1) The NSC has no emergency shutdown or emergency sump pumps. This standard is not applicable.

2) The NSC does not have emergency coolant sources, and so no surveillance requirement is needed or possible.

3) The SAR has no requirements for in-service inspections. This standard is not applicable.

4) Pool water is analyzed frequently, but there is no TS for this assay at the NSCR. A TS will be added to follow the format and content of ANSI/ANS 15.1-2007, Section 4.3.

5) The NSC operates a large surface area non-enclosed pool in a facility with at least 180,000 cubic feet of free space, so the concentration of hydrogen is negligible. This standard is not applicable.

6) While the NSC strongly believes that pH measurements of low conductivity water are suspect, a pH surveillance requirement will be added to 14.4.4.2.

14.4.4 Coolant Systems 14.4.4.1 Coolant Radioactivity Analysis Applicability This specification applies to the surveillance requirements for coolant radioactivity analysis.

Objective The objective is to know the radioactive content of the pool water.

Specifications A sample of the coolant shall be collected and analyzed for radioactive contamination at least weekly during periods of operation and at least quarterly when not operating.

Basis Weekly sampling during operation will allow the establishment of trends of radioactive contamination from fuel or other sources.

14.4.4.2 Coolant Conductivity and pH Applicability This specification applies to the surveillance requirements for coolant conductivity..

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be measured and recorded weekly.

2. pH of the bulk pool water shall be measured and recorded quarterly.

Basis A small rate of corrosion continuously occurs in any water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity from neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the coolant system, which will limit/minimize personnel exposure during maintenance and operations.

An LCO on radioactivity in pool water is added to 3.8.1 Primary Coolant Purity:

14.3.8.1 Primary CoolantPurity Applicability This specification applies to the quality of the primary coolant in contact with the fuel cladding.

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be no higher than 5 x 10-6 mhos/cm (5 pSiemens/cm).

2. The pH of the bulk pool water shall be between 5.5 and 7.5.

3. The concentrations of radionuclides in the bulk pool water shall be no higher than the values presented for water in 10 CFR Appendix B to Part 20 Table 2.

Basis A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is

required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations.

RAI 29

The TS states that the ventilation system shall be verified operable weekly. While the NSC can operate a filter bank, it is not a part of our design basis and therefore not a part of our TS requirement. The filter bank is normally not inline. In fact, we must activate a bypass line from the central exhaust to bring the filter on line. It is not an integral part of the ventilation system.

RAI 3O A tech spec will be added to include a description of the facility location , as well as, the location of exclusion and restricted areas:

The Texas A&M University Nuclear Science Center is located on a rectangular six-acre site on the Texas A&M University campus 1,500 feet from the North-South runway of Easterwood Airport. The facility is six miles south of the city-center of Bryan (pop. 65,660), 3 miles southwest of the main campus of Texas A&M and two and one-half miles west-southwest of the city of College Station (pop. 67,890) in Brazos County, Texas. The NSC is surrounded by land owned and controlled by Texas A&M University and Easterwood Airport. A chain-link steel fence, which provides reasonable restriction of access to the site, defines the confines of the site. This fence also serves to define both the exclusion and restricted areas of the facility. The main entrance into the site is through an electrically operated chain-link steel gate at the east end of the site. Located within the boundaries of the site are the reactor confinement building, reception room, laboratory building, mechanical equipment room, cooling system equipment, holding tanks, and other storage and support buildings. The attached Diagram showing the site boundary, which defines the exclusion and restricted areas will be added to the SAR.

RAI 31

The clarification is that for maintenance procedures that require the reactor to be unsecured, maintenance personnel may replace the Reactor Operator or Reactor Operator Trainee. For example, a control rod inspection requires a control rod to be removed from the core. During this removal the reactor is unsecured, and so the staffing requirement of TAMU TS 14.6.1.3(1) is in force. Rather than requiring a Senior Reactor Operator, a Reactor Operator or Reactor Operator Trainee, AND maintenance personnel to be on site, the note allows only a Senior Reactor Operator and one other person to be on site. TS 14.6.1.3 will be modified such that the unsecured reactor is always under the direct control of the Senior Reactor Operator while he and a maintenance technician are satisfying staffing requirements during maintenance.

14.6.1.3 Staffing Note: During periods of reactor maintenance the reactor operator or the operator trainee may be replaced by maintenance personnel. The licensed senior reactor operator may be permitted as the only operations person present at the facility to perform a pre-startup check of the reactor or perform general reactor maintenance. The unsecured reactor shall always be under the direct control of the Senior Reactor Operator.

RAI 32

TS 14.6.1.4 Selection and Training of Personnel will be updated to endorse the appropriate sections of ANS 15.4 for selection and training of personnel and requalification of licensed operators.

14.6.1.4 Selection and Training of Personnel The selection and training of operations personnel shall be in accordance with the following:

1. Responsibility: The NSC Director or his designated alternate is responsible for the training and requalification of the facility reactor operators and senior reactor operators.

2. Training Program: The Training Program shall be consistent with the standards related to training in ANSI/ANS-1 5.4-2007.

3. Requalification Program: The Requalification Program shall be consistent with the standards related to requalification in ANSI/ANS-15.4-2007.

RAI 33

We will add the following to TS 14.6.2.2:

14.6.2.2 RSB Charter and Rules The operations of the RSB shall be in accordance with a written charter, including provisions for:

1. Meeting Frequency: The RSB shall meet annually at intervals not to exceed 15 months. (Note: The facility license requires a meeting at least once per year and as frequently as circumstances warrant consistent with effective monitoring of facility activities).

2. Quorum: A quorum is comprised of 3 voting members of the RSB excluding ex-officio members.

3. Voting Rules: On matters requiring a vote, if only a quorum is present a unanimous vote of the quorum is required; otherwise a majority vote is required.

4. Subcommittees: The Chairman may appoint subcommittees comprised of one or more members of the RSB including ex-officio members to perform certain tasks.

The subcommittees may be authorized to act for the board.

5. Meeting Minutes: The Chairman will designate one individual to act as recording secretary. It will be the responsibility of the secretary to prepare the minutes which will be distributed to the RSB. The RSB will review and approve the minutes of the previous meetings. A complete file of the meeting minutes will be maintained by the Chairman of the RSB and by the Director of the NSC.

RSB 34 Maintaining exposures and releases as low as reasonably achievable (ALARA) will be added to TAMU TS 14.6.3.

1.1.2 14.6.3 Procedures Written operating procedures shall be prepared, reviewed, and approved before initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the NSC Director or his designated alternate, the Reactor Safety Board, and shall be documented in a timely manner. Procedures shall be adequate to ensure the safe operation of the reactor but shall not preclude the use of independent judgment and action should the situation require such.

Operating procedures shall be in effect for the following items:

1. Startup, operation, and shutdown of the reactor;

2. Fuel loading, unloading, and movement within the reactor;

3. Control rod removal or replacement;

4. Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance of major components of systems that could have an effect on reactor safety;

5. Surveillance checks, calibrations, and inspections of reactor instrumentation and controls, control rod drives, area radiation monitors, facility air monitors, the ventilation system and other systems as required by the Technical Specifications ;

6. Administrative controls for operations, maintenance, and conduct of irradiations and experiments, that could affect reactor safety or core reactivity;

7. Civil disturbances on or near the facility site;

8. Implementation of required plans such as emergency or security plans;

9. Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactivity changes;

10. Radiation protection program to maintain exposures and releases as low as reasonably achievable (ALARA); and

11. Surveillance procedures for shipping radioactive materials.

RAI 35

TS 14.6.3 Procedures will be updated to read:

14.6.3 Procedures Written operating procedures shall be prepared, reviewed, and approved before initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the NSC Director or his designated alternate, the Reactor Safety Board, and shall be documented in a timely manner. Procedures shall be adequate to ensure the safe operation of the reactor but shall not preclude the use of independent judgment and action should the situation require such.

Operating procedures shall be in effect for the following items:

1. Startup, operation, and shutdown of the reactor;

2. Fuel loading, unloading, and movement within the reactor;

3. Control rod removal or replacement;

4. Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance of major components of systems that could have an effect on reactor safety;

5. Surveillance checks, calibrations, and inspections of reactor instrumentation and controls, control rod drives, area radiation monitors, facility air monitors, the ventilation system and other systems as required by the Technical Specifications ;

6. Administrative controls for operations, maintenance, and conduct of irradiations and experiments, that could affect reactor safety or core reactivity;

7. Civil disturbances on or near the facility site;

8. Implementation of required plans such as emergency or security plans;

9. Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactivity changes;

10. Radiation protection program to maintain exposures and releases as low as reasonably achievable (ALARA); and

11. Surveillance procedures for shipping radioactive materials.

RAI 36

TAMU TS 14.6.6.2(1)(a) will be modified to include the words "that includes a discussion of the effect of the violation upon the health and safety of the public."

RAI 37

TAMU TS 14.6.7 will be modified to include these three conditions.

14.6.7.4 Records to be Retained until the Commission terminates the license of the reactor

1. Exceeding a safety limit.

2. Failure of an automatic safety function to function as required.

3. Failure to meet a Limiting Condition for Operation.

14 TECHNICAL SPECIFICATIONS Included in this section are the technical specifications and the bases for the technical specifications. These bases, which provide the technical support for the individual technical specifications, are included for informational purposes only. They are not part of technical specifications and they do not constitute limitations or requirements to which the licensee must adhere.

14.1 Definitions 14.1.1 30/20 LEU Fuel The individual unirradiated LEU fuel elements shall have the following characteristics:

1. Uranium content: maximum of 30 Wt% enriched to maximum 19.95% Uranium-235 with nominal enrichment of 19.75% Uranium-235

2. Hydrogen-to-zirconium atom ratio (in the ZrHx): nominal 1.6 H atoms to 1.0 Zr atoms with a maximum H to Zr ratio of 1.65

3. Natural erbium content (homogeneously distributed): nominal 0.90 Wt%.

4. Cladding: .

14.1.2 ALARA The ALARA program (As Low as Reasonably Achievable) is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environs as low as reasonably achievable.

14.1.3 Annually Items which are designated by these specifications or by standard operating procedures to be performed annually should be performed at least once every 12 months, but at no time shall the time period between consecutive performances of these items exceed 15 months.

14.1.4 Audit An audit is a quantitative examination of records, procedures, or other documents after implementation from which appropriate recommendations are made.

14.1.5 Biennially Items which are designated by these specifications or by standard operating procedures to be performed biennially should be performed at least once every 24 months, but at no time shall the time period between consecutive performances of these items exceed 30 months.

14.1.6 Channel A channel is the combination of sensors, lines, amplifiers, and output devices, which measure the value of a parameter.

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

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

14.1.6.3 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 measuring the same variable.

14.1.7 Confinement Confinement means a closure of the overall facility that results in the control of the movement of air into and out of the facility through a defined path.

14.1.8 Control Rod A control rod is a device fabricated from neutron-absorbing material or fuel, or both, that is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod can be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

14.1.8.1 Regulating ControlRod The regulating rod is a low-worth control rod used primarily to maintain an intended power level that need not have scram capability and may have a fueled follower. Its axial position may be varied manually or by the servo controller.

14.1.8.2 Shim Safety ControlRod A shim safety rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled follower section.

14.1.8.3 TransientControlRod The transient rod is a control rod with scram capabilities that is capable of providing rapid reactivity insertion to produce a pulse.

14.1.9 Core Configuration The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and regulating/shim-safety/transient rods occupying the core grid.

14.1.10 Core Lattice Position The core lattice position is that region in the core (approximately 3" x 3") over a grid-plug hole.

A fuel bundle, an experiment, or a reflector element may occupy the position.

14.1.11 Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff = 1) at reference core conditions.

14.1.12 Experiment An operation, hardware, or target (excluding devices such as detectors, foils, samples etc.) which is designed to investigate non-routine reactor characteristics, or which is intended for irradiation within the pool, on or in a beam port or irradiation facility, and which is not rigidly secured to a core or shield structure so as to be a part of their design.

14.1.12.1 SecuredExperiment A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, other forces that are normal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctions.

14.1.12.2 UnsecuredExperiment An unsecured experiment is any experiment or component of an experiment that does not meet the definition of a secured experiment.

14.1.12.3 Movable Experiment A movable experiment is one where it is intended 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.

14.1.13 Experimental Facilities Experimental facilities shall mean beam ports, including extension tubes with shields, thermal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems, and in-pool irradiation facilities.

14.1.14 Experiment Safety Systems Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information for operator intervention.

14.1.15 Fuel Bundle A fuel bundle is a cluster of two, three, or four elements and/or non-fueled elements secured in a square array by a top handle and a bottom grid plate adapter. Non-fueled elements shall be fabricated from stainless steel, aluminum, boron, or graphite materials.

14.1.16 Fuel Element A fuel element is a single TRIGA fuel rod of LEU type.

14.1.17 Instrumented Fuel Element (IFE)

An instrumented fuel element is a special fuel element in which one or more sheathed chromel-alumel or equivalent thermocouples are embedded in the fuel near the horizontal center plane of the fuel element at a point approximately 0.3 inch from the center of the fuel body.

14.1.18 License The written authorization, by the responsible authority, for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or facility requiring licensing.

14.1.19 Licensee An individual or organization holding a license.

14.1.20 LEU Core A LEU core is an arrangement of TRIGA-LEU fuel in a reactor grid plate.

14.1.21 Limiting Safety System Setting (LSSS)

The limiting safety system setting is a setting for automatic protective devices related to those variables having significant safety functions.

14.1.22 Measuring Channel A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.

14.1.23 Measured Value A measured value is the value of a parameter as it appears on the output of a measuring channel.

14.1.24 Operable Operable means a component or system is capable of performing its intended function.

14.1.25 Operating Operating means a component or system is performing its required function.

14.1.26 Operational Core - Steady State A steady state operational core shall be an LEU core which meets the requirements of the Technical Specifications.

14.1.27 Operational Core - Pulse A pulse operational core is a steady state operational core for which the maximum allowable pulse reactivity insertion has been determined.

14.1.28 Operator See Reactor Operator.

14.1.29 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.

14.1.30 Pulse Mode Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.

14.1.31 Quarterly Items that are designated by these specifications or by standard operating procedures to be performed quarterly should be performed at least once for every three months, and at no time should the time period between consecutive performances of these items exceed four months.

14.1.32 Reactivity Worth of an Experiment The reactivity worth of an experiment is the value of the reactivity change that results from the experiment being inserted into or removed from its intended position.

14.1.33 Reactor Console Secured The reactor console is secured whenever all scrammable rods have been verified to be fully inserted and the console key has been removed from the console.

14.1.34 Reactor Operating The reactor is operating whenever it is not secured or shutdown.

14.1.35 Reactor Operator An individual who is licensed to manipulate the controls of a reactor.

14.1.36 Reactor Safety Systems 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.

14.1.37 Reactor Secured A reactor is secured when:

1. It contains insufficient fissile material or moderator present in the reactor and adjacent experiments to attain criticality under optimum available conditions of moderation and reflection, or

2. The reactor console is secured, and

a. No work is in progress involving core fuel, core structure, core reflector material, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and

b. No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value of one dollar.

14.1.38 Reactor Shutdown The reactor is shut down if it is subcritical by at least $1.00 in the reference core condition with the reactivity worth of all installed experiments included and the following conditions exist:

(a) No work is in progress involving core fuel, core structure, installed control rods,

or control rod drives unless the control rod drives are physically decoupled from the control rods; (b) No experiments are moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment, or $1.00, whichever is smaller.

14.1.39 Reference Core Condition The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (<0.30 dollar).

14.1.40 Reportable Occurrence A reportable occurrence is any of the following:

1. Operation with actual safety system settings for required systems less conservative than the limiting safety-system settings specified in the Technical Specifications 14.2.2.

2. Operation in violation of limiting conditions for operation established in the technical specifications.

3. A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or condition is discovered during maintenance, then no report is required.

4. An unanticipated or uncontrolled change in reactivity greater than one dollar.

5. Abnormal and significant degradation in reactor fuel and/or cladding, coolant boundary, or containment boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment.

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

7. A release of radioactive particles into the environment above allowable limits.

14.1.41 Review A review is a qualitative examination of records, procedures or other documents prior to implementation from which appropriate recommendations are made.

14.1.42 Safety Channel A safety channel is a measuring channel in the reactor safety system.

14.1.43 Safety Limit Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. For the Texas A&M NSC TRIGA reactor the safety limit is the maximum fuel element temperature that can be permitted with confidence that no damage to any fuel element cladding will result.

14.1.44 Scram Time Scram time is the elapsed time between the initiation of a scram signal and the instant that the slowest scrammable control rod reaches its fully inserted position.

14.1.45 Semiannually Items which are designated by these specifications or by standard operating procedures to be performed semiannually should be performed at least once every six month, but at no time should the time period between consecutive performances of these items exceed seven and one-half months.

14.1.46 Senior Reactor Operator An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.

14.1.47 Shall, Should and May The word "shall" is used to denote a requirement: the word "should" to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.

14.1.48 Shutdown Margin Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating condition. This margin is determined assuming that the most reactive scrammable rod and any non-scrammable rods are fully withdrawn, and that the reactor will remain subcritical by this calculated margin without any further operator action.

14.1.49 Steady State Mode Steady state mode of operation shall mean operation of the reactor with the mode selector switch in the steady state position.

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

14.1.51 Unscheduled Shutdown An unscheduled shutdown is 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 that could adversely affect safe operation. It does not include shutdowns that occur during testing or check out operations.

14.1.52 Weekly Items which are designated by these specifications or by standard operating procedures to be performed weekly shall be performed at intervals not to exceed 10 days.

14.2 Safety Limit and Limiting Safety System Setting 14.2.1 Safety Limit-Fuel Element Temperature Applicability This specification applies to the temperature of the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

Specification The temperature in a stainless steel-clad TRIGA LEU fuel element shall not exceed 2100 OF (1 150"C) under any conditions of operation.

Basis The most important safety limit for a TRIGA reactor is fuel element temperature. This parameter is well suited as a single specification because it can be measured directly with a thermocouple.

A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure if the fuel element temperature exceeds the temperature safety limit. The fuel element temperature and the ratio of hydrogen to zirconium in the fuel-moderator material determine the magnitude of the pressure buildup. The mechanism for the pressure buildup is the dissociation of hydrogen from the zirconium hydride moderator that has been blended with uranium to form the fuel mixture encased within the fuel element cladding.

The temperature safety limit for the LEU fuel element is based on data which indicates that the internal stresses within the fuel element, due to hydrogen pressure from the dissociation of the zirconium hydride, will not result in compromise of the stainless steel cladding if the fuel temperature is not allowed to exceed 2100'F (1150'C) and the fuel element cladding is water cooled.

14.2.2 Limiting Safety System Setting Applicability This specification applies to the scram setting that prevents the safety limit from being reached.

Objective The objective is to prevent the safety limits from being reached.

Specification The limiting safety system setting shall be 975°F (525'C) as measured in an instrumented fuel element (IFE). The IFE shall be located adjacent to the central bundle with the exception of the corner positions.

Basis The limiting safety system setting (LSSS) is a temperature that, if exceeded, will cause a reactor scram to be initiated preventing the safety limit from being exceeded.

The temperature safety limit for LEU fuel is 2100'F (1150'C). Due to various errors in measuring temperature in the core, it is necessary to arrive at a LSSS for the fuel element safety limit that takes into account these measurement errors. One category of error between the true temperature value and the measured temperature value is due to the accuracy of the fuel element channel and any overshoot in reactor power resulting from a reactor transient during steady state mode of operation. Although a lesser contributor to error, a minimum safety margin of 10% was applied on an absolute temperature basis. Adjusting the fuel temperature safety limit to degrees Kelvin, K, and applying a 10% safety margin results in a safety limit reduction of 150'C.

Applying this first margin of safety, the safety setting would be 10000C for LEU. However, to arrive at the final LSSS it is necessary to allow for the difference between the measured temperature value and the peak core temperature, which is a function of the location of the thermocouple within the core. For example, if the thermocouple element were located in the hottest position in the core, the difference between the true and measured temperatures would be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. However, at the NSCR, this core position is not available due to the location of the transient rod. For the NSCR, the location of the instrumented elements is therefore restricted to the positions closest to the central element. Calculations indicate that, for this case, the true temperature at the hottest location in the core will differ from the measured temperature by no more than 40%. When applying this 40% worst case measurement scenario and considering the previously mentioned sources of error between the true and measured values,

a final LSSS temperature of 975'F (525'C) is imposed on operation. Viewed on an absolute temperature scale, K, this represents a 37% safety margin in the LEU safety limit.

In the pulse mode of operation, the above temperature limiting safety system setting will apply.

However, the temperature channel will have no effect on limiting peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, a temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting the "tail" off the energy transient in the event the pulse rod remains stuck in the fully withdrawn position.

14.3 Limiting Conditions for Operation 14.3.1 Reactor Core Parameters 14.3.1.1 Steady State Operation Applicability This specification applies to the energy generated in the reactor during steady state operation.

Objective The objective is to ensure that the fuel temperature safety limit will not be exceeded during steady state operation.

Specifications The reactor power level shall not exceed 1.0 megawatt (MW) during steady state operation.

Basis Calculations show that reactor operations with a pool temperature of 60 C will not risk reaching a fuel temperature greater than the LSSS, or a DNBR of unity for 1MW steady state.

14.3.1.2 Pulse Mode Operation Applicability This specification applies to the peak temperature generated in the fuel as the result of a pulse insertion of reactivity.

Objective The objective is to assure that respective pulsing will not induce damage to the reactor fuel.

Specification

1. The reactivity to be inserted for pulse operation shall not exceed that amount which will produce a peak fuel temperature of 1526*F (830'C). In the pulse mode the pulse rod shall be limited by mechanical means or the rod extension physically shortened so that the reactivity insertion will not inadvertently exceed the maximum value.

Basis TRIGA fuel is fabricated with a nominal hydrogen-to-zirconium ratio of 1.6 for LEU fuel. This yields delta phase zirconium hydride that has high creep strength and undergoes no phase changes at temperatures over I000°C. However, after extensive steady state operation at 1.0 MW the hydrogen will redistribute due to migration from the central high temperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, the instantaneous temperature distribution is such that the highest values occur at the surface of the element and the lowest values occur at the center. The higher temperatures in the outer regions occur in fuel with a hydrogen-to-zirconium ratio that has now substantially increased above the nominal value. This produces hydrogen gas pressures considerably in excess of the expected for ZrH1.6. If the pulse insertion is such that the temperature of the fuel exceeds 874°C, then the pressure will be sufficient to cause expansion of microscopic holes in the fuel that grows with each pulse. The pulsing limit of 830'C is obtained by examining the equilibrium hydrogen pressure of zirconium hydride as a function of temperature. The decrease in temperature from 874°C to 830'C reduces hydrogen pressure by a factor of two, which is an acceptable safety factor. This phenomenon does not alter the safety limit since the total hydrogen in a fuel element does not change. Thus, the pressure exerted on the clad will not be significantly affected by the distribution of hydrogen within the element.

In practice the pulsing limit of 830'C will be translated to a reactivity insertion limit for each specific core. The peaking factors from the thermocouple element to the hottest spot in the core must be calculated for each core configuration that is to be used. Temperature would then be measured for small pulse insertions.

The initial core calibration in 2006 established the maximum allowable pulse insertion to be

$1.91. Any subsequent pulse insertion change shall only be made after core recalibration and following approval by the NSCR staff.

14.3.1.3 Shutdown Margin Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worth of control rods and experiments. They apply for all modes of operation.

Objective The objective is to ensure that the reactor can be shutdown at all times and to ensure that the fuel temperature safety limit will not be exceeded.

Specifications The reactor shall not be operated unless the shutdown margin provided by control rods is greater than $0.25 with:

1. Irradiation facilities and experiments in place and the total worth of all non-secured experiments in their most reactive state,

2. The highest worth control rod and the regulating rod (if not scrammable) fully withdrawn, and

3. The reactor in reference core condition.

Basis The value of the shutdown margin ensures that the reactor can be shut down from any operating condition even if the highest worth control rod should remain in the fully withdrawn position.

Since the regulating rod is not scrammable, its worth is not used in determining the shutdown reactivity.

14.3.1.4 Core ConfigurationLimitation Applicability This specification applies to a LEU core.

Objective The objective is to ensure that the fuel temperature safety limit will not be exceeded due to power peaking effects in full LEU cores and with various experimental facilities installed.

Specifications

1. The TRIGA core assembly shall be LEU.

2. The reactor shall not be taken critical with a core lattice position vacant except for positions on the periphery of the core assembly. Water holes in the inner fuel region shall be limited to single rod positions. Vacant core positions shall contain experiments or an experimental facility to prevent accidental fuel additions to the core.

3. The instrumented fuel element, if serving as the Limiting Safety System sensor, shall be located adjacent to the central bundle with the exception of the corner positions.

Basis

1. Safety and accident analysis were only performed for a TRIGA core using LEU fuel.

2. Vacant core positions containing experiments or an experimental facility will prevent accident fuel additions to the reactor core. They will be permitted only on the periphery of the core or a single rod position to prevent power peaking in regions of high power density.

3.

Reference:

SAR Section 14.2.2 14.3.1.5 ReactorFuel Parameters Applicability This specification applies to all fuel elements.

Objective The objective is to maintain the integrity of the fuel elements' cladding.

Specifications

1. The reactor shall not be operated knowingly with damaged fuel.

2. A fuel element shall be considered damaged and must be removed from the core if:

a. In measuring the transverse bend, the bend exceeds 0.125 inch over the length of the cladding.

b. In measuring the elongation, its length exceeds its original length by 0.125 inch, or

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

Basis The limit of transverse bend has been shown to result in no difficulty in disassembling fuel bundles. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to ensure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow.

14.3.1.6 Maximum Excess Reactivity Applicability This specification applies to the maximum excess reactivity, above reference core condition, which may be loaded into the reactor core at any time.

Objective The objective is to ensure that the core analyzed in the safety analysis report approximates the operational core within reasonable limits.

Specifications The maximum reactivity in excess of reference core condition shall not exceed 5.5% Ak/k

($7.85).

Basis Although maintaining a minimum shutdown margin at all times ensures that the reactor can be shut down, that specification does not address the total reactivity available within the core. This specification, although over-constraining the reactor system, helps ensure that the licensee's operational power densities, fuel temperatures, and temperature peaks are maintained within the evaluated safety limits. The specified excess reactivity allows for power coefficients of reactivity, xenon poisoning, most experiments, and operational flexibility.

Note: General Atomics' (GA) modeling of the NSCR LEU core predicted an excess reactivity of

$6.22. The initial data for the NSC reactor start up with a cold, xenon-free core in 2006 yielded a measured excess reactivity of $7.483. GA's modeling of the PRNC FLIP core had predicted

$6.26, while the experimentally determined measured value was $7.12.

14.3.2 Reactor Control and Safety Systems 14.3.2.1 Reactor ControlSystems Applicability This specification applies to the information that must be available to the reactor operator during reactor operation.

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

Specifications The reactor shall not be operated in the specific mode of operation unless the measuring channels listed in Table 14-1 are operable.

Table 14-1: Measuring Channels Required for Operation Operating.Mode

...Minimum Number Measur:ing ..... el~

• M easuri g,.Channel . ... Operable

.:. .. ... .,.* S.S.

• ,:.** . ..Pulse .

Fuel Element Temperature X X Linear Power Level I X Log Power Level I X Integrated Pulse Power I X Basis Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level monitors ensure that the reactor power level is adequately monitored for both steady state and pulsing modes of operation. The specifications on reactor power level indications are included in this section, since the power level is related to fuel temperature.

14.3.2.2 Reactor Safely Systems andInterlocks Applicability This specification applies to the reactor safety system circuits and interlocks.

Obiective The objective is to specify the minimum number of reactor safety system channels and interlocks that must be operable for safe operation.

Specifications The reactor shall not be operated unless the safety circuits and interlocks described in Table 14-2 are operable, however during periods of maintenance, surveillance, calibration and repair, the reactor may be operated as necessary to calibrate and ensure channel operability provided that all the conditions of the TS are met during the operation.

Table 14-2: Safety Circuits and Interlocks Required for Operation Operatig M'.ode S Number Operable Function . , S.S.> >* Pulse Fuel Element Temperature I Scram _ the LSSS X X High Power Level 2 Scram < 125% X -

Console Scram Button I Scram X X High Power Level Detector Power 2 Scram on loss of supply voltage X Supply Preset Timer I Transient Rod Scram 15 seconds or -

less after pulse Log Power I Prevent withdrawal of Shim Safeties at <4x 10W W X Log Power I Prevent pulsing above I kW - X Transient Rod Position I Prevent application of air unless fully X inserted Shim Safety and Regulating Rod Position I Prevent withdrawal - X Pulse Stop Electro-Mechanical Interlock 1 Prevent application of air unless -

pulse stop is installed Basis

1. The fuel temperature and high power level scrams provide protection to assure that the reactor can be shutdown before the safety limit on fuel element temperature will be exceeded.

2. The manual console scram allows the operator to shut down the system if an unsafe or abnormal condition occurs.

3. In the event of failure of the power supply for a high power level safety detector, operation of the reactor without adequate instrumentation is prevented.

4. The preset timer ensures that the reactor power level will reduce to a low level after pulsing.

5. The interlock to prevent startup of the reactor at power levels less than 4 x 10-3 W, which corresponds to approximately 2 cps, assures that sufficient neutrons are available for proper startup.

6. The interlock to prevent pulsing at powers above 1 kW assures that the magnitude of the pulse will not cause the fuel element temperature safety limits to be exceeded.

7. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing of the reactor in steady state mode.

8. The interlock to prevent the withdrawal of the shim safeties or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive period.

9. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted prevents a reactor pulse of sufficient worth to exceed the temperature safety limit.

14.3.2.3 Minimum Number of Operable Scrammable ControlRods Applicability This specification applies to the minimum number of operable scrammable control rods in the core, where operable is specified in terms of maximum scram time from the instant that any SCRAM signal is initiated.

Obiective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Specification During operation all scammable control rods shall be operable. The scram time measured from the instant a SCRAM signal is initiated to the instant that the slowest scrammable rod reaches its fully inserted position shall not exceed 1.2 seconds. During core manipulations, i.e. core loading and unloading, all installed scrammable control rods shall be operable.

Basis This specification ensures that the reactor will be promptly shutdown when a scram signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.

14.3.3 Confinement 14.3.3.1 Operationsthat Require Confinement Applicability This specification applies to confinement requirements during operation of the reactor and the handling of radioactive materials.

Objective To maintain normal or emergency airflow into and out of the reactor building during operations that produce or could potentially produce airborne radioactivity.

Specification Confinement of the reactor building shall be required during the following operations:

1. Reactor operating Note: For periods of maintenance to the central exhaust fan, entry doors to the reactor building shall remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Handling of radioactive materials with the potential for airborne release.

Basis

1. This 41Ar basis applies during the conduct of those activities defined as reactor operations.

is produced during operation of the reactor in experimental facilities and in the reactor pool; thus, air control within the building and the exhaust system in necessary to maintain proper airborne radiation levels in the reactor building and release levels in the exhaust stack. Other radioactivity releases to the reactor building must be considered during reactor operation, such as fission product release from a leaking fuel element or a release from fixed experiments in or near the core. Shutdown of the central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The handling of radioactive materials can result in the accidental or controlled release of airborne radioactivity to the reactor building environment or direct release to the building exhaust system. In these cases, the control of air into and out of the reactor building is necessary.

14.3.3.2 Equipment to Achieve Confinement Applicability This specification applies to the equipment and controls needed to provide confinement of the reactor building.

Objective The objective is to ensure that a minimum of equipment is in operation to achieve confinement as specified in Section 14.3.3.1 and that the control panel for this equipment is available for normal and emergency situations.

Specifications

1. The minimum equipment required to be in operation to achieve confinement of the reactor building shall be the central exhaust fan.

Note: During periods of maintenance to the central exhaust fan, entry doors to the reactor building shall remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Controls for establishing the operation of the ventilation system during normal and emergency conditions shall be available in the reception room.

Basis

1. Operation of the central exhaust fan will achieve confinement of the reactor building during normal and emergency conditions when the controls for air input are set such that the central exhaust fan capacity remains greater than the amount of air being delivered to the reactor building. The exhaust fan has sufficient capacity to handle extra air intake to the building during momentary opening of doors. Shutdown of the central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The control panel for the ventilation system provides for manual selection of air input to the reactor building and the automatic or manual selection of air removal. The air supply and exhaust systems work together to maintain a small negative pressure in the reactor building. These controls are available in the reception room for accessibility during emergency conditions.

14.3.4 Ventilation System Applicability This specification applies to the operation of the facility ventilation system.

Objective The objective is to ensure that the ventilation system can mitigate the consequences of the possible release of radioactive materials resulting from reactor operation and radioactive material handling.

Specification The reactor shall not be operated unless the following parts of the facility ventilation system are operable:

1) Central Exhaust Fan

2) Intake louvers on the air handlers The reactor may be operated during periods of maintenance up to one hour. During this maintenance period the reactor building doors shall remain closed except for momentary opening for personnel entry or exit.

The ventilation system shall be secured automatically by signals from the stack particulate, fission product, or stack gas (xenon) facility air monitor.

Basis During normal operation of the ventilation system, the concentration of 41Ar in unrestricted areas is below the effluent concentration (SAR Chapter 11). In the event of a substantial release of airborne radioactivity, the ventilation system will be secured automatically. Therefore, operation of the reactor with the ventilation system shutdown for short periods of time to make repairs ensures the same degree of control of release of radioactive materials. Moreover, facility air monitors within the building independent of those in the ventilation system will give warning of high levels of radiation that might occur during operation with the ventilation system secured.

The ventilation system must be shutdown for weekly operability check. The one hour limit provides enough time to complete these checks.

14.3.5 Radiation Monitoring Systems and Effluents 14.3.5.1 RadiationMonitoring Applicability This specification applies to the radiation monitoring information that must be available to the reactor operator during reactor operation.

Objective The objective is to ensure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.

Specification The reactor shall not be operated unless the radiation monitoring channels listed in Table 14-3 are operable, displays and alarms are operable in the control room, and displays are operable in the ESC.

Table 14-3: Radiation Monitoring Channels Required for Operation RadiationMonitoring Channels Function Number*

Monitor radiation levels within Reactor Bridge ARMvI the reactor bay Stack Particulate Monitor Monitor radiation levels in the (FAM Ch. 1) exhaust air stack Stack Gas Monitor Monitor radiation levels in the (FAM Ch. 3) exhaust air stack Building Particulate Monitor Monitor radiation levels within (FAM Ch. 4) the reactor bay Note: When required monitors are inoperable, or for periods of maintenance, the intent of this specification will be satisfied if it is replaced for a period of no more than 1 week with a portable gamma sensitive instrument having its own alarm which shall be kept under visual observation.

If two of the above monitors are not operating, the reactor shall be shutdown.

Basis The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.,

14.3.5.2 Argon-41 DischargeLimit Applicability This specification applies to the concentration of Argon-41 (41Ar) that may be discharged from the TRIGA reactor facility.

Objective The objective is to ensure that the health and safety of the public is not endangered by the discharge of 41Ar from the TRIGA reactor facility.

Specification The concentration of 4'Ar in the effluent gas from the facility as diluted by atmospheric air in the lee of the facility due to the turbulent wake effect shall not exceed 1.0 x 10-8 ýtCi/ml averaged over one year.

Basis The maximum allowable concentration of 4 1Ar in air in unrestricted areas as specified in Appendix B, Table II of 10CFR20 is 1.0 x 10-8 [tCi/ml. Section 11 of the SAR for the NSCR substantiates a 5.0 x 10-3 atmospheric dilution factor for a 2.0 mph wind speed. This dilution factor represents the conditions at the site building for a wind speed of 2.0 mph, which occurs less than 10% of the time on an annual basis.

14.3.5.3 Xenon andIodine Monitoring Applicability This specification applies to the radiation monitoring systems necessary to monitor and control the concentration of any effluent releases during the production of 1251 from the radioactive decay of 125Xe.

Objective The objective is to ensure sufficient radiation monitoring information is available to the operator, and air handlers are automatically shutdown to ensure that the health and safety of the general public is not endangered during the production of 1251.

Specification No experiment that involves active handling of '2 5Xe and 125I may be performed unless the following radiation monitorig system is operable (Table 14-4). No experiment may be performed, except decay of 12 Xe, unless the 125Xe FAM channel is operable.

Table 14-4: Radiation Monitoring Required for Xenon-125 and Iodine-125 Experiments Radiation Monitoring Channels Number 25 1 Xe Effluent Monitoring Channel (FAM Ch. 5) 1 1251Air Monitor I Note: When required monitors are inoperable, or for periods of maintenance, the intent of this specification will be satisfied if this monitor is replaced for a period of no more than 1 week with a portable gamma sensitive instrument having its own alarm which shall be kept under visual observation.

Basis The required facility air monitors (14.3.5.1) are not calibrated for 125I. The 2' 5Xe FAM channel provides information to operators in the event there is a significant release of 12 5Xe during the production of 125. The 125I air monitor measures 1251 radiation levels in the immediate area where the iodine handling is being performed.

14.3.6 Limitations on Experiments 14.3.6.1 Reactivity Limits Applicability This specification applies to the reactivity limits on experiments installed in the reactor and its experimental facilities.

Objective The objective is to ensure control of the reactor during the handling of experiments adjacent to or in the reactor core.

Specifications The reactor shall not be operated unless the following conditions governing experiments exist.

1. The reactivity worth of any single, movable or unsecured experiment shall be less than one dollar.

2. The reactivity worth of any secured experiment shall be less than two dollars.

3. The sum of the absolute reactivity of all experiments shall be less than five dollars.

Basis

1. This specification is intended to ensure that the worth of a single unsecured experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were suddenly inserted. This does not restrict the number of unsecured experiments adjacent to or in the reactor core except by worth.

2. The maximum worth of a single experiment is limited so that its removal from the reference core condition reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be secured, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

3. This limit poses a restriction on the number of experiments being run at any given time to prevent excessive positive and negative reactivity loading.

14.3.6.2 MaterialLimitations Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

Objective The objective is to prevent damage to the reactor or excessive release of radioactivity by limiting materials quantity and radioactive material inventory of the experiment.

Specifications

1. Explosive materials in quantities greater than 5 pounds (TNT-equivalent) shall not be allowed within the reactor building. Irradiation of explosive materials shall be restricted as follows:

a. Explosive materials in quantities greater than 25 milligrams (TNT-equivalent) shall not be irradiated in the reactor pool. Explosive materials in quantities less than 25 milligrams (TNT-equivalent) may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container.

b. Explosive materials in quantities greater than 25 milligrams (TNT-equivalent) shall be restricted from the reactor pool, the upper research level, the demineralizer room, cooling equipment room and the interior of the pool containment structure.

c. Explosive materials in quantities greater than 5 pounds (TNT-equivalent) shall not be irradiated in experimental facilities.

d. Cumulative exposures for explosive materials in quantities greater than 25 milligrams (TNT-equivalent) shall not exceed 101 n/cm 2 for neutron or 25 Roentgen for gamma exposures.

2. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 10 Ci.

3. Significant amounts of corrosive materials used in a reactor experiment shall be double encapsulated. Exceptions may only be made if a detailed analysis and/or prototype testing with small amounts of materials demonstrates that the experiment presents negligible risk.

Basis

1. This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving explosive materials.

a. This specification is intended to prevent damage to the reactor core and safety related reactor components located within the reactor pool in the event of failure of an experiment involving the irradiation of explosive materials. Limited quantities of less than 25 milligrams (TNT-equivalent) and proper containment of such experiment provide the required safety for in-pool irradiation provided that the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container.

b. This specification is intended to prevent damage to vital equipment by restricting the quantity and location of explosive materials within the reactor building.

Explosives in quantities exceeding 25 milligrams (TNT-equivalent) are restricted from areas containing the reactor bridge, reactor console, pool water coolant and purification systems, and reactor safety related equipment.

c. The failure of an experiment involving the irradiation of up to 5 pounds (TNT-equivalent) of explosive material in an experimental facility located external to the reactor pool structure will not result in damage to the reactor or the reactor pool containment structure.

d. This specification is intended to prevent any increase in the sensitivity of explosive materials due to radiation damage during exposures.

2. The 10 Ci limitation on Iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR 20 for an unrestricted area.

3. This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving corrosive materials.

14.3.6.3 FailuresandMalfunctions Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

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

Specifications

1. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (a) normal operating conditions of the experiment or reactor, (b) credible accident conditions in the reactor, or (c) possible accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor building or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B of 10CFR20.

2. In calculations pursuant to a) above, the following assumptions shall be used:

a. If the effluent from an experimental facility exhausts through a holdup tank that closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles at least 10% of these vapors can escape.

c. For materials whose boiling point is above 130'F and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape.

3. If a capsule fails and releases material that could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the NSC Director or his designated alternate and determined to be satisfactory before operation of the reactor is resumed.

Basis

1. This specification is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10 CFR 20 will be released to the atmosphere outside the facility boundary of the NSC.

2. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment failure.

3. Operation of the reactor with reactor fuel or structure damage is prohibited to avoid release of fission products. Potential damage to reactor fuel or structure must be brought to the attention of the NSC Director or his designated alternate for review to assure safe operation of the reactor.

14.3.6.4 Xenon Irradiationfor Iodine Production Applicability This specification applies to the experiments that produce 125I from the activation of enriched 124Xe and the decay of 125Xe.

Objective The objective is to prevent excessive release of radioactivity by limiting the quantity and radioactive material inventory of the experiment.

Specifications

1. 124Xe activation experiments shall be controlled such that the total single experiment activity produced is limited to no more than 2000 Ci of 125Xe.

2. The total facility 125Xe inventory of all experiments shall not exceed 3500 Ci.

Basis

1. The 2000 Ci limitation on Xenon-125 produced in any one experiment assures that in the event of a failure of an experiment leading to the accidental release of xenon, the exposure to the general public would be less than 0.05 rem per year (10 CFR 20.1301).

2. In the NRC's safety evaluation supporting amendment 15 to our facility license the commission concluded that the NSC's request for a total inventory limit of Xe-125 of 3500 Ci was reasonable based on the facility experimental design and operating plan for Iodine Production.

14.3.7 As Low As Reasonable Achievable (ALARA) Radioactive Effluents Released Applicability This specification applies to the measures required to ensure that the radioactive effluents released from the facility are in accordance with ALARA criteria.

Obiective The objective is to limit the annual radiation exposure to the general public resulting from operation of the reactor to a level as low are reasonably achievable below the limits listed in 10 CFR 20.1301.

Specifications

1. In addition to the radiation monitoring specified in Section 14.5.4, an environmental radiation monitoring program shall be conducted to measure the integrated radiation exposure in and around the environs of the facility on a quarterly basis.

2. The annual radiation exposure (dose) to the public due to reactor operation shall not exceed the limits defined in 10 CFR 20.1301. The facility perimeter shall be monitored to ensure this specification is being met.

3. The total annual discharge of 4aAr into the environment shall not exceed 30 Ci per year.

4. In the event of a significant fission product leak from a fuel rod or a significant airborne radioactive release from a sample being irradiated, as detected by the facility air monitor (FAM), the reactor shall be shut down until the source of the leak is located and eliminated. However, the reactor may continue to be operated on a short-term basis, as needed, to assist in determining the source of the leakage.

5. Before discharge, the facility liquid effluents collected in the holdup tanks shall be analyzed for the nature and concentration of radioactive effluents. The total quantity of radioactive material that the is released into the sanitary sewerage system in a year does not exceed 1 curie (37 GBq) of all radioactive materials combined.

Basis The simplest and most reliable method of ensuring that ALARA release limits are accomplishing their objective of minimal facility-caused radiation exposure to the general public is to actually measure the integrated radiation exposure in the environment on and off the site.

14.3.8 Primary Coolant Conditions 14.3.8.1 PrimaryCoolant Purity Applicability This specification applies to the quality of the primary coolant in contact with the fuel cladding.

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be no higher than 5 x 10.6 mhos/cm (5 jtSiemens/cm).

2. The pH of the bulk pool water shall be between 5.5 and 7.5.

3. The concentrations of radionuclides in the bulk pool water shall be no higher than the values presented for water in 10 CFR Appendix B to Part 20 Table 2.

Basis A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations.

14.3.8.2 PrimaryCoolant Level and Leak Detection Applicability This specification applies to the physical level of primary coolant that must be in the pool and requirements for leak detection for reactor operation.

Objective The objective is to ensure proper shielding and cooling of the reactor and the ability to detect leaks.

Specification

1) The reactor shall not be operated if the pool level is lower than 18 feet above the top of the core.

2) The reactor shall not be operated if the pool level alarm is not operable, and capable of producing an alarm at a constantly monitored off-site facility.

Basis

1) The intake to the diffuser system is approximately 18 feet above the core. Setting this level as the specification will both ensure the availability of the diffuser system, and provide more than adequate shielding and cooling for the reactor.

2) An operable pool level alarm that notifies off-site will ensure proper notification if a low pool level is reached.

14.3.8.3 PrimaryCoolant Temperature Applicability This specification applies to the maximum allowable primary coolant temperature.

Objective The objective is to maintain fuel temperature less than the LSSS, and to maintain the departure of nucleate boiling ratio (DNBR) greater than unity, and to limit any degradation of the reactor systems.

Specification

1) The Director, or his designee, shall be notified when pool temperature reaches 50 C, with further operation dependent on his approval.

2) The reactor shall not be operated when pool temperature exceeds 60 C.

Basis Calculations show that reactor operations with a pool temperature of 60 C will not risk reaching a fuel temperature greater than the LSSS, or a DNBR of unity. For reactor pulses, the NSC already accounts for pool temperature. As described in the SAR (4.5.13 Pulse Operation - NSC

- BOL, Measured), the NSC calculates peak core temperature. The temperature in this calculation is in °C. Ambient fuel temperature is already included in pulse calculations.

14.4 Surveillance Requirements 14.4.1 General Applicability This specification, applies to the surveillance requirements of any system related to reactor safety.

Obiective The objective is to verify the proper operation of any system related to reactor safety.

Specifications Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications

approved by the Reactor Safety Board. A system shall not be considered operable until it is successfully tested.

Basis This specification relates to changes in reactor systems, which could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.

14.4.2 Reactor Core Parameters 14.4.2.1 Steady State Operation Applicability This specification applies to the surveillance requirement of the power level monitoring channels.

Objective The objective is to verify that the maximum power level of the reactor meets the license requirements.

Specification A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually.

Basis The power level channel calibration will assure that the reactor will be operated at the proper power level.

14.4.2.2 Pulse Mode Operation Applicability This specification applies to the surveillance requirements for operation of the reactor in the pulse mode.

Obiective The objective is to verify that operation of the reactor in the pulse mode is proper and safe and to determine if any significant changes in fuel characteristics have occurred.

Specification The reactor shall be pulsed semiannually to compare fuel temperature measurements and core pulse energy with those of previous pulses of the same reactivity value. The reactor shall not be declared operational for pulsing until such pulse measurements are performed.

Basis The reactor is pulsed at suitable intervals to make a comparison with previous similar pulses and to determine if changes in fuel or core characteristics are taking place.

14.4.2.3 Shutdown Margin Applicability This specification applies to the surveillance requirement of control rod calibrations and shutdown margin.

Objective The objective is to verify that the requirements for shutdown margins are met for operational cores.

Specification The reactivity worth of each control rod and the shutdown margin shall be determined annually and following significant core configuration and/or control rod changes.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to insure no significant changes in the shutdown margin.

14.4.2.4 Excess Reactivity Applicability This specification applies to the surveillance requirements of reactor excess reactivity.

Objective The objective is to verify that requirements on excess reactivity are met for operational cores.

Specification The excess reactivity shall be determined biennially and following significant core configuration and/or control rod changes.

Basis The excess reactivity of the core is measured to ensure that during all states of operation criticality can be maintained for licensed operational limits. With the accumulation of fission product poison buildup and fissile material bumup, excess reactivity must be available for power transients and maintaining criticality.

14.4.2.5 Reactor Fuel Elements Applicability This specification applies to the surveillance requirements for the fuel elements.

Objective The objective is to verify the continuing integrity of the fuel element cladding and to ensure that no fuel damage has occurred.

Specification

1. The following fuel elements shall be inspected visually for damage or deterioration and measured for length and bend annually.

a. At least four elements which occupy the highest pulse temperature positions in the core;

b. At least one-fifth of the fuel elements used in operation of the reactor over the previous inspection year;

c. The four elements (a) above may be included in the inspection of the fuel elements of (b) above.

2. If any element is found to be damaged, the entire core will be inspected.

Basis The frequency of inspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known. Experience has shown that temperature is the major contributor to fuel damage. Inspection of four fuel elements which occupy the highest pulse temperature positions in the core provides surveillance for detection of the most probable

fuel element damage should it occur. Inspection of one-fifth of the elements used in operation of the reactor provides surveillance of the lower temperature elements and over a five year period provides for inspection of all elements.

14.4.3 Reactor Control and Safety Systems 14.4.3.1 Reactor ControlSystem Applicability These specifications apply to the surveillance requirements for reactor control systems.

Objective The objective is to verify the condition and operability of system components affecting safe and proper control of the reactor.

Specifications

1. The control rods shall be visually inspected for deterioration biennially.

2. Operability tests of the control rod mechanism shall follow modification or repairs.

3. The Transient Rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated semiannually.

Basis

1. The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation of the reactor.

2. These tests provide verification that the control rod has full travel and that the rod drop time is within specification.

3. Inspection and maintenance of the transient rod drive assembly reduces the probability of failure of the system due to moisture-induced corrosion of the pulse cylinder and piston rod assembly.

14.4.3.2 Reactor Safety Systems Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of the control and safety systems.

Objective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.

Specifications

1. A channel check of each of the reactor safety system channels for the intended mode of operation shall be performed before each day's operation or before each operation extending more than one day, except for the pool level alarm channel, which shall be tested weekly.

2. Whenever a reactor scram caused by high power level or high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety limit was exceeded.

3. A calibration of the fuel element temperature measuring channels shall be performed semiannually.

4. A calibration of the pool temperature measuring channel shall be performed semiannually.

5. A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated.

6. Reactor safety systems which include interlocks shall be tested at least annually.

Basis Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. If the period between operations extends beyond a year, then the annual channel test requirement will assure operability.

14.4.3.3 Scram Time Applicability This specification applies to the surveillance of control rod scram times.

Objective The objective is to verify that all scrammable control rods meet the scram time requirement.

Specification The scram time shall be measured annually.

Basis Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.

14.4.4 Coolant Systems 14.4.4.1 Coolant RadioactivityAnalysis Applicability This specification applies to the surveillance requirements for coolant radioactivity analysis.

Obiective The objective is to know the radioactive content of the pool water.

Specifications A sample of the coolant shall be collected and analyzed for radioactive contamination at least weekly during periods of operation and at least quarterly when not operating.

Basis Weekly sampling during operation will allow the establishment of trends of radioactive contamination from fuel or other sources.

14.4.4.2 Coolant Conductivity andpH Applicability This specification applies to the surveillance requirements for coolant conductivity.

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be measured and recorded weekly.

2. pH of the bulk pool water shall be measured and recorded quarterly.

Basis A small rate of corrosion continuously occurs in any water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system

is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity from neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the coolant system, which will limit/minimize personnel exposure during maintenance and operations.

14.4.5 Equipment to Achieve Confinement: Ventilation System Applicability This specification applies to the building confinement ventilation system.

Objective The objective is to ensure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment.

Specification During periods of operation, or radioactive material handling, the ventilation system shall be verified operable weekly. This specification is not required during periods of non-operation, e.g.,

holidays, extended maintenance outages.

Basis Experience accumulated over several years of operation has demonstrated that the tests of the ventilation system on a weekly basis are sufficient to assure the proper operation of the system and control of the release of radioactive material.

14.4.6 Radiation Monitoring Systems and Effluents Applicability This specification applies to the surveillance requirements for the area radiation monitoring equipment and the facility air monitoring (FAM) system.

Objective The objective is to ensure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.

Specification The area radiation monitoring system (ARM) and the facility air monitoring system (FAM) shall be calibrated annually and shall be tested weekly.

Basis Experience has shown that weekly verification of area radiation and air monitoring system operations in conjunction with annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span.

14.4.7 Experiments Applicability This specification applies to the surveillance requirements for experiments installed in the reactor and its experimental facilities and for irradiations performed in the irradiation facilities.

Objective The objective is to prevent the conduct of experiments or irradiations that may damage the reactor or release excessive amounts of radioactive materials as a result of failure.

Specifications

1. A new experiment shall not be installed in the reactor or its experimental facilities until a hazard analysis has been performed and reviewed for compliance with Section 14.6 of the Technical Specifications. Minor modifications to a reviewed and approved experiment may be made at the discretion of the Director, or his designee, with concurrence from the Radiation Safety Officer, or his designee. The Director, or his designee, and the Radiation Safety Officer, or his designee, shall review the hazards associated with the modifications and determine that the modifications do not create a significantly different, a new, or a greater safety risk than the original approved experiment, and does not require a review under 10CFR50.59.

2. The performance of an experiment classified as an approved experiment shall not be performed until a licensed senior operator and the Radiation Safety Officer, or his designee has reviewed it for compliance.

3. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before reactor operation with said experiment.

Basis It has been demonstrated over a number of years of experience that experiments and irradiations reviewed by the Reactor Staff and the Reactor Safety Board as appropriate can be conducted without endangering the safety of the reactor or exceeding the limits in the technical specifications.

14.4.8 Surveillance Deferment Applicability This specification describes which surveillance requirements may be deferred during reactor shutdown, and which shall be performed prior to resuming reactor operations.

Objective The objective is to give facility personnel the ability to avoid conducting unnecessary surveillances during periods of reactor shutdown.

Specifications Possible to Defer TS During Shutdowns? Required prior to operations?

I. 14.4.2.1 Steady State Operation Yes Yes

2. 14.4.2.2 Pulse Mode Operation Yes Yes

3. 14.4.2.3 Shutdown Margin Yes Yes

4. 14.4.2.4 Excess Reactivity Yes Yes

5. 14.4.2.5 Reactor Fuel Elements No N/A

6. 14.4.3.1 Reactor Control Systems No N/A

7. 14.4.3.2 Reactor Safety Systems Yes, except pool level Yes

8. 14.4.3.3 Scram Time Yes Yes

9. 14.4.4.1 Coolant Radioactivity Analysis Yes - Quarterly Yes

10. 14.4.4.2 Coolant Conductivity and pH Yes - Quarterly Yes

11. 14.4.5 Equipment to Achieve Confinement:

Ventilation System Yes Yes

12. 14.4.6 Radiation Monitoring Systems and Effluents No N/A

13. 14.4.7 Experiments Yes Yes Basis

1. It is impossible to perform a calorimetric while the reactor is shutdown; a calorimetric shall be performed prior to resuming operations.

2. It is impossible to perform pulse surveillance while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

3. It is impossible to determine control rod worth while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

4. It is impossible to determine control rod worth while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

5. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

6. The possibility for control rod degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

7. With the exception of the pool level alarm, whose surveillance shall not be deferred, it is not necessary to conduct surveillance on reactor safety systems, which are effectively automatic scram systems, while the reactor is shutdown. The surveillance shall be performed prior to resuming operations.

8. It is impossible to determine scram time while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

9. The possibility for cladding degradation or damage still exists while the reactor is shutdown; however it is much less likely. Therefore the periodicity of this surveillance may be changed to quarterly. The surveillance shall be performed prior to resuming operations.

10. Fuel cladding degradation is still undesirable while the reactor is shutdown; however chemically contaminating the pool is much less likely. Therefore the periodicity of this surveillance may be changed to quarterly. The surveillance shall be performed prior to resuming operations.

11. The ventilation system is only required during reactor operations or radioactive materials handling. The surveillance shall be performed prior to resuming operations or handling.

12. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

13. No experiments can be installed in the reactor while it is shutdown; the surveillance shall be performed prior to resuming operations.

14.5 Design Features 14.5.1 Site and Facility Description Applicability This specification applies to the NSC site location and specific facility design features.

Obiective The objective is to specify the location of specific facility design features.

Specifications

1. The restricted area is that area inside the fence surrounding the site.

2. The reactor confinement building houses the TRIGA reactor.

3. The reactor bay shall be equipped with a. ventilation system designed to exhaust air or other gases from the reactor confinement building and release them from a stack at a minimum of 85 feet from ground level.

4. Emergency shutdown controls for the ventilation systems shall be located in the reactor control room and emergency support center.

Basis The restricted area is described in SAR 2.2.1.1. The facility is designed such that the ventilation system will maintain a negative pressure in the reactor confinement building with respect to the outside atmosphere so that there will be no uncontrolled discharge to the environment. Controls for startup and operation of the ventilation system are located in the reactor control room and emergency support center.

14.5.2 Reactor Fuel Applicability This specification applies to the fuel elements used in the reactor core.

Objective The objective is to ensure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications TRIGA LEU 30/20 Fuel: The individual unirradiated LEU fuel elements shall have the following characteristics:

1. Uranium content: maximum of 30 Wt% enriched to maximum 19.95% Uranium-235 with nominal enrichment of 19.75% Uranium-235

2. Hydrogen-to-zirconium atom ratio (in the ZrHx): nominal 1.6 H atoms to 1.0 Zr atoms with a maximum H to Zr ratio of 1.65

3. Natural erbium content (homogeneously distributed): nominal 0.90 Wt%.

4. Cladding: .

Basis The fuel specification permits a maximum uranium enrichment of 19.95%. This is about 1%

greater than the design value for 19.75% enrichment. Such an increase in loading would result in an increase in power density of less than 1%. An increase in local power density of 1% reduces the safety margin by less than 2%.

The fuel specification for a single fuel element permits a minimum erbium content of about 5.6%

less than the design value of 0.90 wt%. (However, the quantity of erbium in the full core must not deviate from the design value by more than -3.3%). This variation for a single fuel element would result in an increase in fuel element power density of about 1-2%. Such a small increase in local power density would reduce the safety margin by less than 2%.

The maximum hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of 2 greater than for a hydrogen-to-zirconium ratio of 1.60. This increase in the clad stress during an accident would not exceed the rupture strength of the clad.

14.5.3 Reactor Core Applicability This specification applies to the configuration of fuel and in core experiments.

Objective The objective is to ensure that provisions are made to restrict the arrangement of fuel elements and experiments to provide assurance that excessive power densities will not be produced.

Specifications

1. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator bundles positioned in the reactor grid plate.

2. The reflector, excluding experiments and experimental facilities, shall be any combination of graphite, water, and heavy water.

Basis

1. Standard TRIGA cores have been in use for years and their characteristics are well documented. LEU cores including 30/20 fuel have also been operated at General Atomics

and their successful operational characteristics are available. General Atomics and Texas A&M have conducted a series of studies documenting the viability of using LEU fuel in TRIGA reactors.

2. The core will be assembled in the reactor grid plate that is located in a pool of light water.

Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.

14.5.4 Control Rods Applicability This specification applies to the control rods used in the reactor core.

Objective The objective is to ensure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications

1. The shim safety control rods shall have scram capability and contain borated graphite, B 4C powder, or boron and its compounds in solid form as a poison in aluminum or stainless steel cladding. These rods may incorporate fueled followers that have the same characteristics as the fuel region in which they are used.

2. The regulating control rod may not have scram capability and shall be a stainless rod or contain the materials as specified for shim safety control rods. This rod may incorporate a fueled follower.

3. The transient control rod shall have scram capability and contain borated graphite or boroni and its compounds in solid form as a poison in an aluminum or stainless steel clad.

The transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an aluminum or air follower.

Basis Using neutron absorbing borated graphite, B4C powder, or boron and its compounds, satisfies the poison requirements for the control rods. Since the regulating rod normally is a low worth rod, using a solid stainless steel rod could satisfy its function. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to ensure mechanical stability during movement and to isolate the poison from the pool water environment. Control rods that are fuel followed provide additional reactivity to the core and increase the worth of the control rod. The use of fueled followers in the LEU fueled region has the additional advantage of reducing flux peaking in the water filled regions vacated by the withdrawal of the control rods. Scram capabilities are provided for rapid insertion of the control rods, which is the primary safety

feature of the reactor. The transient control rod is designed for a reactor pulse. The nuclear behavior of the air or aluminum follower that is incorporated into the transient rod is similar to a void. A voided follower may be required in certain core loadings to reduce flux peaking values.

14.5.5 Radiation Monitoring System Applicability This specification describes the functions and essential components of the area radiation monitoring (ARM) equipment and the facility air monitoring (FAM) system for continuously monitoring airborne radioactivity.

Objective The objective is to describe the radiation monitoring equipment available to the operator to assure safe operation of the reactor.

Specification The radiation monitoring equipment listed in Table 14-5 shall have the following characteristics:

Table 14-5: NSC Radiation Monitoring Equipment Radiation Monitoring Dete.tor Type.. . Function Channel Area ARatRadiation Monitor Monitor radiation fields in key locations. Alarm Gamma sensitive detector and readout in the control room and readout in the reception room.

Facility Air Monitor (FAM) - Monitors concentration of airborne radioactive Particulates Beta-Gamma sensitive detector particulates. Alarm and readout in the control room and readout in the reception room.

Monitors Alarm andconcentration readout in theofcontrol radioactive Facility Air Monitor (FAM) - Gamma sensitive detector room gases.

and readout in the reception room.

Basis The radiation monitoring system is intended to provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.

14.5.6 Fuel Storage Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The objective is to ensure that fuel that is being stored will not become critical and will not reach an unsafe temperature.

Specifications

1. All fuel elements and fueled devices shall be stored in a geometrical array for which the k-effective is less than 0.8 for all conditions of moderation and reflection.

2. Irradiated fuel elements and fueled devices shall be stored in an array, which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.

Basis The limits imposed by Specifications 14.5.5.1 and 14.5.5.2 are conservative and ensure safe storage.

14.5.7 Reactor Building Ventilation System Applicability This specification applies to the building that houses the reactor.

Objective The objective is to ensure that provisions are made to restrict the amount of release of radioactivity into the environment.

Specifications

1. The reactor shall be housed in a facility designed to restrict leakage. The minimum free volume in the facility shall be 180,000 cubic feet.

2. The reactor building shall be equipped with a ventilation system designed to exhaust air or other gases from the reactor building and release them from a stack at a minimum of 85 feet from ground level.

3. Emergency shutdown controls for the ventilation system shall be located in the reception room and the system shall be designed to shut down in the event of an alarm on the fission product facility air monitor.

Basis The facility is designed such that the ventilation system will normally maintain a negative pressure with respect to the atmosphere so that there will be no significant uncontrolled leakage

to the environment. The free air volume within the reactor building is confined when there is an emergency shutdown of the ventilation system. Controls for startup, emergency filtering, and normal operation of the ventilation system are located in the reception room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reception room minimizing exposure to operating personnel.

14.5.8 Reactor Pool Water Systems Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.

Objective The objective is to ensure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.

Specifications

1. The reactor core shall be cooled by natural convective water flow.

2. The pool water inlet and outlet pipe for the demineralizer, diffuser, and skimmer systems shall not extend more than 15 feet below the top of the reactor pool when fuel is in the core.

3. Pool water inlet and outlet pipes to the heat exchanger shall have emergency covers within the reactor pool for manual shut off in case of pool water loss due to external pipe system failure.

4. A pool level alarm shall indicate loss of primary coolant before or equal to the pool level dropping to 10% below the normal operating level (28.8 feet of water in the pool).

Basis

1. This specification is based on thermal and hydraulic calculations, which show that the TRIGA-LEU core can operate continuously in a safe manner at power levels up to 2,420 kW, with natural convection flow and sufficient bulk pool cooling.

2. In the event of accidental siphoning of pool water through inlet and outlet pipes of the demineralizer, skimmer, or diffuser systems, the pool water level will drop to no more than 15 feet from the top of the pool, providing 17 feet of water above the core.

3. Inlet and outlet coolant lines to the pool heat exchanger terminate at the bottom of the pool. In the event of pipe failure, these lines must be manually sealed from within the reactor pool. Covers for these lines will be stored in the reactor pool. The time required to

uncover the reactor core due to failure of a single pool coolant pipe system

4. Coolant water loss of 10% or more requires corrective action. This alarm is observed in the reactor control room, in the reception room, and at a continuously manned remote location.

14.6 Administrative Controls 14.6.1 Organization The Nuclear Science Center is operated by the Texas A&M University System's Texas Engineering Experiment Station (TEES), with responsibility within TEES resting with the Director or his designee. The Director of the Nuclear Science Center is responsible to the TEES licensee for the administration and the proper and safe operation of the facility. Figure 12-1 shows the administration chart for the Nuclear Science Center. The Reactor Safety Board advises the director of the NSC on all matters or policy pertaining to safety. The NSC Radiological Safety Officer provides "onsite" advice concerning personnel and radiological safety and provides technical assistance and review in the area of radiation protection.

Figure 14-1: Organization Chart for Reactor Administration

14.6.1.1 Structure

1. A line management organizational structure provides administration and operation of the reactor facility.

2. The TEES licensee and the director of the NSC have line management responsibility for adhering to the terms and conditions of the Nuclear Science Center Reactor (NSCR) license and technical specifications and for safeguarding the public and facility personnel from undue radiation exposure. The facility shall be under the direct control of the NSC Director or a licensed senior reactor operator.

3. Management Levels:

a. Level 1: TEES Licensee (Director of TEES): Responsible for the NSCR facility license.

b. Level 2: NSC Director: Responsible for reactor facility operation and shall report to Level 1.

c. Level 3: Senior Reactor Operator on Duty: Responsible for the day-to-day operation of the NSCR or shift operation and shall report to Level 2.

d. Level 4: Reactor Operating Staff: Licensed reactor operators and senior reactor operators and trainees. These individuals shall report to Level 3.

4. Radiation Safety:

A qualified health physicist has the responsibility for implementation of the radiation protection program at the NSCR. The individual reports to Level 2 management.

5. Reactor Safety Board (RSB):

The RSB is responsible to the licensee for providing an independent review and audit of the safety aspects of the NSCR.

14.6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be in accordance with the line organization established in Section 14.6.1.1.

14.6.1.3 Staffing

1. The minimum staffing when the reactor is not secured shall be as follows:

a. At least two individuals shall be present at the facility complex and shall consist of at least a licensed senior reactor operator and either a licensed reactor operator or operator trainee.

Note: During periods of reactor maintenance the reactor operator or the operator trainee may be replaced by maintenance personnel. The licensed senior reactor operator may be permitted as the only operations person present at the facility to perform a pre-startup check of the reactor or perform general reactor maintenance. The unsecured reactor shall always be under the direct control of the Senior Reactor Operator.

b. A licensed reactor operator or senior reactor operator shall be in the control room.

c. The NSC Director or his designated alternate is readily available for emergencies or on call (i.e., capable of getting to the reactor facility within 30 minutes).

d. At least one member of the health physics support group shall be readily available at the facility or on call (i.e., capable of getting to the reactor facility within 30 minutes), to provide advice and technical assistance in the area of radiation protection.

2. A list of reactor facility personnel by name and telephone number shall be readily available for use in the control room. The list shall include:

a. Administrative personnel

b. Radiation safety personnel

c. Other operations personnel

3. The following designated individuals shall direct the events listed:

a. The NSC Director or his designated alternate shall direct any loading of fuel or control rods within the reactor core region.

b. The NSC Director or his designated alternate shall direct any loading of an in-core experiment with a reactivity worth greater than one dollar.

c. The senior reactor operator on duty shall direct the recovery from an unplanned or unscheduled shutdown other than a safety limit violation.

d. The senior reactor operator on duty shall direct the initial startup and approach to power during each operating day.

14.6.1.4 Selection and Trainingof Personnel The selection and training of operations personnel shall be in accordance with the following:

1. Responsibility: The NSC Director or his designated alternate is responsible for the training and requalification of the facility reactor operators and senior reactor operators.

2. Training Program: The Training Program shall be consistent with the standards related to training in ANSI/ANS- 15.4-2007.

3. Requalification Program: The Requalification Program shall be consistent with the standards related to requalification in ANSI/ANS-15.4-2007.

14.6.2 Review and Audit 14.6.2.1 ReactorSafety Board(RSB)

The Reactor Safety Board shall be comprised of at least 3 voting members knowledgeable in fields which relate to Nuclear Safety. One of these members, normally the Deputy Director of TEES, will serve as the Chairman. If the Chairman is unable to attend one or a number of committee meetings he may designate a committee member as Chairman Pro-tem. The members are normally appointed by the licensee to serve one year terms. It is expected that the members will be reappointed each year as long as they are willing to serve so that their experience and familiarity with the past history of the NSC will not be lost to the committee. The Director of the NSC, TAMU Radiological Safety Officer, Head of the Department of Nuclear Engineering, and a senior member of the NSCR Radiation Protection Office shall be ex-officio members of the RSB.

14.6.2.2 RSB Charterand Rules The operations of the RSB shall be in accordance with a written charter, including provisions for:

1. Meeting Frequency: The RSB shall meet annually at intervals not to exceed 15 months. (Note: The facility license requires a meeting at least once per year and as frequently as circumstances warrant consistent with effective monitoring of facility activities).

2. Quorum: A quorum is comprised of 3 voting members of the RSB excluding ex-officio members.

3. Voting Rules: On matters requiring a vote, if only a quorum is present a unanimous vote of the quorum is required; otherwise a majority vote is required.

4. Subcommittees: The Chairman may appoint subcommittees comprised of one or more members of the RSB including ex-officio members to perform certain tasks.

The subcommittees may be authorized to act for the board.

5. Meeting Minutes: The Chairman will designate one individual to act as recording secretary. It will be the responsibility of the secretary to prepare the minutes which will be distributed to the RSB. The RSB will review and approve the minutes of the previous meetings. A complete file of the meeting minutes will be maintained by the Chairman of the RSB and by the Director of the NSC.

14.6.2.3 RSB Review Function The review responsibilities of the Reactor Safety Board or a designated subcommittee shall include, but are not limited to the following:

1. Review and evaluation of whether a proposed change, test, experiment, or changes in equipment and systems would constitute an unevaluated 10CFR50.59 safety question or a change in technical specification;

2. Review of new procedures, major revisions of procedures, and proposed changes in reactor facility equipment or systems which have significant safety impact to reactor operations;

3. Review of new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity;

4. Review of proposed changes to the technical specifications, license or charter;

5. Review of the NSCR radiation protection program;

6. Review of violations of technical specifications, license, or charter, and violations of internal procedures or instructions having safety significance;

7. Review of operating abnormalities having safety significance;

8. Review of reportable occurrences listed in 14.6.6.2;

9. Review of audit reports 14.6.2.4 RSB Audit Function The RSB or a subcommittee thereof shall audit reactor operations and radiation protection programs quarterly. Audits shall include but are not limited to the following:

1. Facility operations, including radiation protection, for conformance to the technical specifications, applicable license conditions, and standard operating procedures annually;

2. The results of action taken to correct those deficiencies that may occur in the reactor facility equipment systems, structures, or methods of operations that affect reactor safety annually;

3. The retraining and requalification program for the operating staff annually;

4. The facility security plan and records annually;

5. The reactor facility emergency plan and implementing procedures annually.

14.6.2.5 Audit of ALARA Program The licensee or his designated alternate (excluding anyone whose normal job function is within the NSCR) shall conduct an audit of the reactor facility ALARA program annually. The licensee shall transmit the results of the audit to the RSB at the next scheduled meeting.

14.6.3 Procedures Written operating procedures shall be prepared, reviewed, and approved before initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the NSC Director or his designated alternate, the Reactor Safety Board, and shall be documented in a timely manner. Procedures shall be adequate to ensure the safe operation of the reactor but shall not preclude the use of independent judgment and action should the situation require such.

Operating procedures shall be in effect for the following items:

1. Startup, operation, and shutdown of the reactor;

2. Fuel loading, unloading, and movement within the reactor;

3. Control rod removal or replacement;

4. Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance of major components of systems that could have an effect on reactor safety;

5. Surveillance checks, calibrations, and inspections of reactor instrumentation and controls, control rod drives, area radiation monitors, facility air monitors, the ventilation system and other systems as required by the Technical Specifications ;

6. Administrative controls for operations, maintenance, and conduct of irradiations and experiments, that could affect reactor safety or core reactivity;

7. Civil disturbances on or near the facility site;

8. Implementation of required plans such as emergency or security plans;

9. Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactivity changes;

10. Radiation protection program to maintain exposures and releases as low as reasonably achievable (ALARA); and

11. Surveillance procedures for shipping radioactive materials.

14.6.4 Experiment Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures.

1. All new experiments or class of experiments shall be reviewed by the RSB (14.6.2.3) and implementation approved in writing by the NSC Director or his designated alternate.

2. Substantive changes to previously approved experiments shall be made only after review by the RSB and implementation approved in writing by the NSC Director or his designated alternate. The NSC Director or his designated alternate may approve minor changes that do not significantly alter the experiment.

14.6.5 Required Actions 14.6.5.1 Action to be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2. An immediate report of the occurrence shall be made to the Chairman, RSB, NSC Director, and reports shall be made to the NRC in accordance with Section 14.6.6.2 of these specifications, and

3. A report shall be prepared which shall include an analysis of the cause and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the RSB for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

14.6.5.2 Action to be Taken in the Event of a Reportable Occurrence In the event of a reportable occurrence, the following action shall be taken:

1. NSC staff shall return the reactor to normal operating or shut down conditions. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the NSC Director or his designated alternate.

2. The NSC Director or his designated alternate shall be notified and corrective action taken with respect to the operations involved.

3. The Director or his designated alternate shall notify the Chairman of the RSB.

4. A report shall be made to the RSB which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.

5. A report shall be made to the NRC in accordance with Section 14.6.6.2 of these specifications.

6. Occurrence shall be reviewed by the RSB at their next scheduled meeting.

14.6.6 Reporting Requirements 14.6.6.1 Annual Report An annual report covering the operation of the reactor facility during the previous calendar year shall be submitted to the NRC before March 31 of each year providing the following information:

1. A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections;

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

3. The number of emergency shutdowns and inadvertent scrams, including reasons thereof;

4. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required;

5. A brief description, including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50;

6. 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 before the point of such release or discharge. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.

a. Liquid Waste (summarized on a monthly basis)

i. Radioactivity discharged during the reporting period.

1. Total radioactivity released (in Curies).

2. The effluent concentration used and the isotopic composition if greater than 1x 10-7 O tCi/cc for fission and activation products.

3. Total radioactivity (in curies), released by nuclide during the reporting period based on representative isotopic analysis.

4. Average concentration at point of release (in [tCi/cc) during the reporting period.

ii. Total volume (in gallons) of effluent water (including dilution) during periods of release.

b. Airborne Waste (summarized on a monthly basis)

i. Radioactivity discharged during the reporting period (in Curies) for:

l. 41/kAr

2. Particulates with half-lives greater than eight days.

c. Solid Waste

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

ii. The total activity involved (in Curies).

iii. The dates of shipment and disposition (if shipped off site).

7. A summary of radiation exposures received by facility personnel and visitors, including dates and time where such exposures are greater than 25% of that allowed or recommended.

8. A description and summary of any environmental surveys performed outside the facility.

14.6.6.2 Special Reports In addition to the requirements of applicable regulations, reports shall be made to the NRC Document Control Desk and special telephone reports of events should be made to the Operations Center as follows:

1. There shall be a report not later than the following working day by telephone and confirmed in writing by fax or similar conveyance to be followed by a written report that

describes the circumstances of the event and includes a discussion of the effect of the violation upon the health and safety of the public within 14 days of any of the following:

a. Violation of safety limits (14.6.5.1).

b. Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure;

c. Any reportable occurrences as defined in 14.1.40. The written report (and, to the extent possible, the preliminary telephone or faxed report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of the event;

2. A written report within 30 days of:

a. Personnel changes in the facility organization involving Level 1 and Level 2.

b. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.

14.6.7 Records Records of facility operations in the form of logs, data sheets, or other suitable forms shall be retained for the period indicated as follows:

14.6.7.1 Records to be Retainedfor a Periodof at Least Five Years orfor the Life of the Component Involved

1. Normal Reactor facility operation.

2. Principal maintenance operations.

3. Reportable occurrences.

4. Surveillance activities required by the technical specifications.

5. Reactor facility radiation and contamination surveys where required by applicable regulations.

6. Experiments performed with the Reactor.

7. Fuel inventories, receipts, and shipments.

8. Approved changes in operating procedures.

9. Records of meeting and audit reports of the RSB.

14.6.7.2 Records to be Retainedfor at Least One Training Cycle

1. Retraining and Requalification of certified operations personnel.

2. Records of the most recent complete cycle shall be maintained for individuals employed.

14.6.7.3 Records to be Retainedfor the Lifetime ofthe Reactor Facility

1. Gaseous and liquid radioactive effluents released to the environs.

2. Off-site environmental monitoring surveys required by the technical specifications.

3. Radiation exposure for all personnel monitored.

4. Drawings of the reactor facility.

14.6.7.4 Records to be Retained untilthe Commission terminates the license of the reactor

1. Exceeding a safety limit.

2. Failure of an automatic safety function to function as required.

3. Failure to meet a Limiting Condition for Operation.